ML030790039
| ML030790039 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/10/2003 |
| From: | Gallagher M AmerGen Energy Co |
| To: | Document Control Desk, Office of Nuclear Security and Incident Response |
| References | |
| 2130-03-20057 NUMARC/NESP-007, Rev 2 | |
| Download: ML030790039 (138) | |
Text
AmerGen Energy Company, LLC wwwexeloncorp corn AmerGenM An Exelon/British Energy Company 2oo Exelon Way Suite 345 Kennett Square, PA 19348
.". *.10CFR50
..... Appendix E.IV.B 2130-03-20057 March 10, 2003 U. S. Nuclear Regulatory Commission Attn: Document Control Desk "
Washington, DC 20555
Subject:
Oyster Creek Generating Station Implementation of Emergency Action Levels Developed from NUMARC/NESP-007 Methodology Oyster Creek Generating Station Facility License No. DPR-16 Docket No. 50-219 "
Enclosed for your review and'apjroval are new Emergency Action Levels (EALs) and the associated Technical Bases Manual that is proposed for implementation at Oyster Creek Generating Statiori. These EALs are based on the methodology outlined in NUMARCINESP-007, "Methodology for Development of Emergancy A-tion Levels," which is endorsed in Revision 3 to Regulatory Guide 1.101, "Emergency Planning and Preparedness for Nuclear Power Plants,"
dated August 1992.
The Oyster Creek Generating Station Plant Operations Review Committee (PORC) has reviewed and approved the package for submittal to the NRC. -NRC approval is required in.
accordance with the provisions of 10 CFR 50, Appendix E.IV.B.
In order to support the NRC's review of the Technical Bases Manuals, the following information is enclosed.
Oyster Creek Generating Station Upgraded EAL Technical Bases Manual Oyster Creek Generating Station EAL Table Oyster Creek Generating Station Upgraded EAL NUMARC Comparison The enclosed EAL Technical Bases Manual and supporting documentation have been reviewed and agreed upon with the State of New Jersey, Bureau of Nuclear Engineering. A letter indicating the State's acceptance is-attached. These EALs.will not be implemented until .
completion of the NRC's review and approval, and until initial training is completed.
Changes to the Nuclear Emergency Plan and procedures resulting from implementation of revised EALs will be performed in accordance with the requirements of 10 CFR 50.54(q),
subsequent to NRC approval of this change.
A64f 5
2130-03-20057 March 10, 2003 Page 2 We are requesting that the NRC review and approve the EALs contained within the Technical Bases Manual for Oyster Creek Generating Station which incorporate the NUMARC/NESP-007 guidelines by August 29, 2003, in order to facilitate training of the appropriate Emergency Response Organization (ERO) by the end of October 2003.
If you should have any questions, please contact Mr. John G. Hufnagel at 610-765-5507.
Very truly yours, Michael P. Gallagher Director, Licensing and Regulatory Affairs AmerGen Energy Company, LLC Enclosures cc: H. J. Miller, USNRC Administrator, Region I P. S. Tam, USNRC Senior Project Manager, Oyster Creek R. J. Summers, USNRC Senior Resident Inspector, Oyster Creek File No. 03005
98475i3;* 2 3-10-03; 1:27PM 3--10--03 ; 1 :27PM ; 9847513;t 2 James E. McGreevey Department of Environmental Protection Govwrmr Bradley M Campbell Divbiion ofEnvironmental Safety, Health, and Analytical Programs Commiwioner Radiation Protetion Programs Bureau of Nucloar Engineering PO Box 415 TrCnton, New Jcrwy 0862.5-0415 Phone- (609) 984-7709 March 10, 2003 Mr. Ernest Harkness, Vice President AmerGen Energy Company Oyster Creek Nuclear Generating Station Route 9 South P.O. Box 388 Forked River, N-J. 08731-0388 Dear Mr. Harkness The New Jersey Bureau of Nuclear Engineering has completed the review of the proposed Emergency Action Levels (EALs) that AmerGen is planning to submit to the NRC for approval. Over the past several months AmerGen and the Bureau of Nuclear Engineering have participated in working meetings to review the proposed Oyster Creek Nuclear Generating Station EALs, EAL Bases Document, and the EAL Matrix developed from NUMARC/NESP-007 Methodology.
The revised EALs, which incorporate the resolution of our comments and discussions, are acceptable to the State of New Jersey. If you should have any questions, please feel free to contact me at 609-984-7700.
Sin erely, Kent Tosch aager Bureau of Nuclear Engineering New Jersey is an Equal OpporiunityEmployer ROCYCL'eidPaper
Summary of Differences NUMARC/NESP-007 Rev. 2 to Proposed OCNS Emergency Action Levels NUMARC/
NESP-007 NExPle7 OCNS EAL 0OCNS EAL Differences/Justification Example #
IC/EAL AULI. JRU None AU1.2 RU1 None 3N/A OCNS does not have telemetered perimeters monitors, therefore this EAL is not required AU1.4 N/A OCNS does not use automatic initiation of real time dose assessment therefore this EAL is not required AU2.1 RU6 None AU2.2 RU5 None AU2.3 RU7 None AU2.4 RU2 None AAI.l RAI None AA1.2 RA1 None OCNS does not have telemetered perimeters monitors therefore this EAL is not required OCNS does not use automatic initiation of real time dose assessment therefore this EAL is not required AA2.1 RA5 None AA2.2 RA5 None AA2.3 RA6 None AA2.4 RA5 None AA3.1 RA2 None AA3.2 RA2 None ASlI.1 RS1 None AS 1.2 NIA OCNS does not have telemetered perimeters monitors therefore this EAL is not required AS 1.3 RS1 None AS 1.4 RS1 None AGI.1 RG1 None AG 1.2 N/A OCNS does not have telemetered perimeters monitors therefore this EAL is not required AG1.3 RG1 None AG1.4 RG1 None HUI.1 HU3 None HU1.2 HU3 None HU1.3 HU3 None HU1.4 HU3 None Unanticipated explosions covered under HU4 (NUMARC HU1.5 HU3 HU2) since explosions are more logically associated with OCNS threshold HU4 HU1.6 HU3 None 1
Summary of Differences NUMARC/NESP-007 Rev. 2 to Proposed OCNS Emergency Action Levels NUMARC/
NESP-007 OCNSEAL OCNS EAL Differences/Justification Example #
IC/EAL Added site specific high and low intake water levels as other conditions appropriate for this IC HU2 HU4 Included unanticipated explosions from HU1.5 HU3.1 HU5 None HU3.2 HU5 None This EAL threshold has been written to conform with IC HU4 HU4.1 HUI regarding devices as amended and endorsed by the NRC in a letter from Mr. B. A. Boger to Ms. Lynette Hendricks (NEI) dated 2/4/02.
HU4.2 HUI This wording conforms to the criteria of HU4 as amended and approved by NRC for post-9/1 1 security issue resolutions HU5 HU6 None OCNS does not have installed seismic instrumentation to determine if seismic activity is in excess of OBE levels.
Procedure 2000-ABN-3200.38 "Station Seismic Event" HAl.1 HA3 requires the Shift Manager to scram the reactor for conditions in which the seismic activity causes a threat to safe plant operation. This is consistent with earthquakes in excess of OBE levels and consistent with the existing OCNS seismic analysis.
HA1.2 HA3 None HA1.3 HA3 None HA1.4 HA3 None HA1.5 HA3 None HA1.6 N/A No plant safety equipment is potentially impacted by missiles generated by turbine rotating failures at OCNS.
HA1.7 HA 3 Added site specific high and low intake water levels as other conditions appropriate for this IC HA2 HA4 None HA3.1 HA5 None HA3.2 HA5 None HA4.1 HAI None HA4.2 HA1 None HA5 HA2 None HA6 HA6 None HS1.L HSI None HS1.2 HSI None HS2 HS2 None HS3 HS6 None HG1.1 HGI None HG1.2 HG1 None 2
Summary of Differences NUMARC/NESP-007 Rev. 2 to Proposed OCNS Emergency Action Levels NUMARC/
NESP-007 NExPle07 OCNSEAL Example # OCNS EAL Differences/Justification IC/EAL HG2 HG6 None SUl MU1 None SU2 MU9 None The condition "In the opinion of the Shift Supervisor, the loss of the annunciators or indicators requires increased SU3 M'U6 surveillance to safely operate the unit(s)" has not been included in the condition consistent with changes to IC SU3 in NUMARC/NESP-007 Rev. 3. This statement does not provide useful assessment criteria to the EAL threshold.
SU4.1 RU4 None The MODE applicability [1,2] is a deviation from NUMARC SU4.2 RU3 [all] in that, the SJAE Radiation Monitor, selected as an 'other indication' will only be a valid indication of Fuel Clad Degradation mode's [1, 2].
"Pressure boundary leakage" not applicable to OCNS since no SU5 MU7 distinction is made between unidentified or pressure boundary leakage in the OCNS Technical Specifications.
SU6 MU8 None SU7 MU3 None SA1 MA2 None Added "Loss of manual SCRAM capability indicated by failure of ALL manual SCRAM attempts to achieve reactor shutdown" per resolution of NJ BNE concerns and consintentcy with Hope Creek Station.
SA3 MA5 None The condition "In the opinion of the Shift Supervisor, the loss of the annunciators or indicators requires increased SA4 AMl6 surveillance to safely operate the unit(s)" has not been included in the condition consistent with changes to IC SU3 in NUMARC/NESP-007 Rev. 3. This statement does not provide useful assessment criteria to the EAL threshold.
SA5 MA1 None SS1 MS1 None SS2 MS4 None SS3 MS3 None Implements NUMARC/NESP-007 Rev. 3 BWR specific criteria. Revision 2 of NUMARC/NESP-007 simply specified loss of [site-specific function] necessary to maintain Hot Shutdown. Revision 3 of NUMARC is specific in defining this condition for BWRs as inability to maintain parameters below Heat Capacity Temperature Limit.
3
Summary of Differences NUMARC/NESP-007 Rev. 2 to Proposed OCNS Emergency Action Levels NUMARC/
NESP-007 OCNS EAL OCNS EAL Differences/Justification Example #
IC/EAL __
The condition stated in NUMARC NESP-007, SS5, L.a "Loss of all decay heat removal cooling as determined by (site specific) procedure" is not necessary to conclude that the plant condition warrants a Site Area Emergency due to core uncovery; therefore, the example EAL was not included in this SS5 MS7 EAL.
Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level.
RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
SS6 MS6 None Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level.
SG1 MG1 RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
SG2.1 MG4 None Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level.
SG2.2 MG4 RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
FC.1 J.d.1 None Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level.
FC.2 1.a.1/2 RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
"AND Indication of RCS leak inside drywell" criteria added to FC.3 .b.] clarify intent and distinguish from loss of containment cooling events which can manifiest itself symptomatically similar to RCS leakage which is the intent of the IC.
FC.4 N/A No 'other' fuel clad loss/potential loss indicators identified for OCNS FC.5 1I None RC. I 2.c.] None Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level.
RC.2 2.a.] RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
4
Summary of Differences NUMARC/NESP-007 Rev. 2 to Proposed OCNS Emergency Action Levels NUMARC/
NESP-007 NExPle7 OCNS EAL 0OCNS EAL Differences/Justification Example #
IC/EAL RC.3 2.d.1/2/3/4 None RC.4 2.b.1 None RC.5 N/A No 'other' RCS loss/potential loss indicators identified for OCNS RC.6 2./.I None PC. 1 c.23 3.e.] - None The Revsion 2 NUMARC example EAL prescribes an RPV water level in conjunction with the Maximum Core Uncovery Time Limit (MCUTL). This is a misapplication of the PC.2 3.a.] MCUTL, which was corrected in revision 3 of NUMARC/NESP-007. Primary Containment Flooding required (Entry into SAMG) is now specified in the current NUMARC document.
PC.3 3.d.1/2/3 None PC.4 3.b.1 None PC.5 N/A No 'other' PC loss/potential loss indicators identified for OCNS PC.6 3f.I None 5
Oyster Creek Nuclear Station Annex Section 3 Emergency Action Levels (EALs)
EAL Technical Bases Revision Oc Prepared By; Operations Support Services, Inc.
1716 White Pond Lane Waxhaw, NC 28173 (704) 243-0501 www.ossi-net.com Prepared For Exelon Nuclear 200 Exelon Way Kennett Square, PA 19348 Purchase Order #01042079
Oyster Creek Nuclear Station Annex Exelon Nuclear Section 3: Classification of Emergencies Section D of the Exelon Nuclear Standardized Radiological Emergency Plan describes five (5) Emergency Classes. The first four are the Unusual Event, Alert, Site Area Emergency and General Emergency, and are listed from least severe to most severe according to relative threat to the health and safety of the public and emergency workers.
The fifth level is Recovery and is considered as a phase of the emergency. Recovery is not considered as part of the event classification logic contained in Section 3.0 of the Annex, but rather is entered by meeting criteria provided in Section M of the Exelon Nuclear Standardized Radiological Emergency Plan.
Site specific definitions are provided for terms to be used for that particular Initiating Conditions /Threshold Values and may not be applicable to other uses of that term in any other EAL, at other sites, in the Exelon Nuclear Standardized Radiological Emergency Plan or procedures. Also included are the technical bases, which were used to develop the EAL.
All classifications are to be based upon VALID indications, reports or conditions.
Indications, reports or conditions are considered VALID when they are verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
When two or more Emergency Action Levels are determined, declaration will be made on the highest classification level for the Unit.
3.1 Emergency Action Levels (EALs)
Emergency Action Levels are the measurable, observable detailed conditions that must be met in order to classify the event. Classification shall not be made without referencing, comparing and satisfying the threshold values specified in the Emergency Action Levels. Mode Applicability provides the unit conditions when the Emergency Action Levels represent a threat. The Basis provides definitions of terms, explanations and justification for including the Initiating Condition and Emergency Action Level. Definitions are provided for terms having specific meaning as they relate to this procedure.
Unusual Event, Alert, Site Area Emergency, and General Emergency classifications are entered by meeting designated Emergency Action Levels (EALs) Threshold Values. These values are based on the criteria established under Revision 2 to NUMARC/NESP-007, "Methodology for Development of Emergency Action Levels" (dated January 1992), and are labeled based on the four Recognition Categories outlined in NUMARC/NESP-007:
"* Abnormal Radiological Levels / Effluents
"* Fission Product Barrier Degradation
"* System Malfunctions Page 2 of 122 Revision 0c
Ovter Crtek Nnc1e2r St2tinn Ann *Y*,lnn Nllrl*nr 0 Hazards and Other Conditions EAL Threshold Values are sorted under common Initiating Conditions (ICs).
These ICs can be Symptom- or Event-based, and applicable to all or only designated Operational Conditions / Modes OPCONs. The Initiating Conditions (IC) and associated EAL Threshold Values are summarized in the EAL Matrix (Table OCNS 3-1) according to Recognition Categories.
To aid user in identifying applicable ICs, they are further sorted under the following Event Sub-Categories, and appropriate Mode designator provided:
" Abnormal RadiologicalLevels /Effluents ("1')
- Radiological Effluents
- Abnormal Rad Levels
- Coolant Activity
- Irradiated Fuel Accidents
" FissionProductBarriers("F")
FissionProductBarrierMatrix comprised of.
- Fuel Clad Barrier
- Reactor Coolant System (RCS) Barrier
- Primary Containment (PC) Barrier
" System Malfunctions ("M')
- Loss of AC Power
- Loss of DC
- Failure of RPS
- Loss of Annunciators
- Loss of Communications
- Technical Specifications Hazards and Other Conditions ("H')
- Security Events
- Control Room Evacuation
- Natural or Man-Made Events
- Fire / Explosion
- Toxic or Flammable Gas
- Discretionary Page 3 of 122 Revision Oc
Oyster Creek Nuclear Station Annex Exelon Nuclear An emergency is classified by assessing plant conditions and comparing abnormal conditions to ICs and Threshold Values for each EAL, based on the designated Operational Condition (MODE). Modes 1 through 4 are based on Reactor Mode Switch Position and average reactor coolant temperature. "Defueled" Mode was established for classification purposes under NUMARC/NESP-007 to reflect conditions where all fuel has been removed from the Reactor Pressure Vessel.
MODE TITLE CONDITION 1 Power Operation Includes Run and Startup Modes when > 212 'F 2 Hot Shutdown ?> 212 OF 3 Cold Shutdown < 212 OF 4 Refueling D Defueled The EAL Matrix is designed to provide an evaluation of the Initiating Conditions from the worst conditions (General Emergencies) on the left to the relatively less severe conditions on the right (Unusual Events). Evaluating conditions from left to right will reduce the possibility that an event will be under classified. All Recognition Categories should be reviewed for applicability prior to classification.
An appropriate EAL numbering system is provided as a user aid. ICs are coded with a two letter and one number code. For example: HA1 The first letter is the Recognition Category designator. In this case, H stands for "Hazards and Other Conditions". The second letter is the Classification Level:
"U" for Unusual Event, "A" for Alert, "S" for Site Area Emergency, and "G" for General Emergency. The number is a sequential number for that Recognition Category series. All Initiating Conditions, which are describing the severity of a common condition (series), will have the same number (e.g. HAl, HA2, etc.).
A Fission Product Barrier (FPB) Table is provided as a subset to the Recognition Category "F" (FPB Degradation) of the EAL Matrix. This table is used to determine the integrity of the Fuel Clad, RCS and Containment Barriers based on EAL Threshold values established in accordance with NUMARC/NESP-007 (e.g., Intact, LOSS, or POTENTIAL LOSS).
Page 4 of 122 Revision 0c
- rolnn Nllrlonr 3.2 EAL Technical Basis Table OCNS 3-2 serves as the Technical Basis for the EAL Matrix. The table consists of the following sections for each Initiating Condition (IC), sorted by Recognition Category:
- Initiating Condition
- Threshold Value(s)
- Mode Applicability
- Basis
- Plant-specific References
- Differences (deviations from NUMARC/NESP-007 as appropriate)
Table OCNS 3-2 provides the EAL user with the background and justification behind the EAL Threshold Values identified using the guidance set forth in NUMARC/NESP-007.
3.3 General EAL Implementation Philosophy A broad spectrum of discretion in classifying events is provided in the "Discretionary" category under Hazards and Other Conditions and the Fission Product Barrier Matrix in Table OCNS 3-1. In using the "Discretionary" category and in classifying emergencies under circumstances which are not a straight forward use of the EAL's, ERO members should be mindful than an approach is needed which is conservative with respect to public, plant, and personnel safety and with respect to ensuring the adequacy of personnel and technical support.
Conservative decisions must be made if the ED has any doubt regarding the health and safety of the public.
Declaring an Unusual Event provides the Company and off-site agencies the opportunity for early information regarding the event and for early activation of resources and may be considered a "no consequence decision." Conversely, not declaring an Unusual Event when there is credible (but, not clear) bases for doing so, would appear to be less than open or candid and could have serious adverse consequences. Although the consequences of declaring an Unusual Event are limited, inappropriate classifications do not accurately indicate the significance of the event to the public and emergency responders and should be avoided.
At the Alert, Site Area and General Emergency levels, clearly the threat to the plant and to the public is at a heightened level. Rapid application of resources and preparation for providing for the public health and safety are appropriate. Because of the magnitude of resource mobilization and the potential disruption of normal public activities, an overly conservative or an inappropriately early declaration of these levels is not advisable.
Page 5 of 122 Revision 0c
Events that meet the Emergency Action Level criteria for event declaration, but which are terminated before they are identified and declared, should still be classified and reported, but not declared to implement the Emergency Plan.
All EALs may not consider trends, rates of change, or status changes in equipment availability. In the event of rapidly changing parameters trending toward an increased emergency classification, it may be appropriate to decide that the higher level EAL will be exceeded and escalate the classification early. In the event of trends toward a decreased emergency classification, parameter values must be below the EALs to de-escalate.
In the event of a "spike" which rapidly exceeds and then decreases below an EAL, entry into the Emergency Plan or escalation to the higher classification "in retrospect" is not appropriate unless the "spike" is indicative of continuing degrading conditions which will lead to an escalated emergency classification level. This statement does not apply if the EAL includes a "spike". Spurious alarms or parameters, which are known to be invalid indicators of actual plant conditions or of the emergency classification, should not be used to declare emergency classifications.
Page 6 of 122 Revision 0c
it'3vtfar '*raolt *,,t.loa.. qfaHnn A.nav IExelon INuclear TABLE. OC.NS 31- E.-mereeny ActUon Level i(.AL, Mlatrilx (conilt GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ABNORMAL RAD LEVELS EFFLUENTS RAS Major Damage OR Uncovering of _____RU5 Potential Damage OR Uncovering of Spent 34D Spent Fuel M Fuel EAL Threshold Valuep FAL Threshold Value, Nonei None 1. Valid unanticipated Hi alarm on one or more Refuel Floor Ii Uncontrolled water level drop in the Spent Fuel Pool with i
- i ARMsOR (Table R-3) ALL irradiated fuel assemblies remaining covered by water 2 Report of visual observation that irradiated fuel in the Spent
_ _ _ _ __ [ Fuel Pool or Fuel Transfer Canal is uncovered RA6 Loss of Wate Level That Has OR Will RU6 Uncontrolled Water Level Drop in Uncover Irradiated Fuel as the Reactor iM Retr C.aviy I Cavity EAL Threshold Value.
EAL Threshold Value,, 1 Unexpected Skimmer Surge Tank Lo-Lo level alarm None None 1. Report of visual observation that irradiated AND fuel in the Reactor Cavity is or will be I Visual observation of an uncontrolled drop in uncovered water level below the fuel pool skimmer surge >
_ _ _ _ _ _tank inlet IRU? Independent Spent Fuel Storage Inst~ilation iEAL Threshold Value, ri L"
II. Radiation readings > 10 times normal at ANY of the following ISFSI locations None None None
__________________________________________ I . on contact with shield wall Table R-3: Refuel Floor ARMs
- C-5, Cnt Mon C 9, North Wall
- C-10, North Wall
- B-9, Open Floor Page 8 of 122 Revision 0c
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Y. e ree ticear on nexExelon Nuclear TABLE PBAPS 3-I: Emergency Action Level (EAT.) Matrix (Cont'd)
GENERAL EMERGENCY -7S1 E AREA EMERGENCY ALLRI UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS (cont.)
HA3 Natural OR Destructive Phenomena iHU3 Natural OR Destructive Phenomena
,Affecting a Vital Area T i Affecting the Protected Area [Ti T1
- EAL Threshold Value, EAL Threshold Value.
- 1. Confirmed earthquake requiring reactor scram in accordance 11. Felt earthquake I with 2000-ABN-3200 38 Station Seismic Event OR OR 12 Report by plant personnel of atomado strike within the 2 Tornado or wind speeds > 100 mph causing damage to Plant Protected Area Vital Structures (Table H-1) OR LT OR 3 Sustained wind speeds> 75 mph as indicated by on-ste
- 3. Report of visible structural damage to ANY Plant Vital meteorological instrumentation Structure (Table H-I) due to natural or destructive phenomena OR None None OR 4 Vehicle crash within the Protected Area Boundary that may 4 Vehicle crash damaging or affecting Plant Vital Structure potentially damage plant structures containing functions and (Table H-I) systems required for safe shutdown of the plant.
- 5. Abnormal Intake Structure level, at indicated by EITHER 5. Report of turbine failure resulting in casing penetration or S >60 ft. MSL (>492 psig on PI-SWS-1[2]) damage to turbine or generator seals OR I OR
_-4.0 ft. MSL (< 050 psg on P6-533-1172 or P-533- lntake Structure level, as indicated by EITHER 1173) >45ft.MSL(>426pstgonPI-SWS-l[2])
OR MSL - Mean Sea Level * -3 0 ft.MSL
(< 0 94 psig on PI-533-1172 or PI-533-1173)
I HA4 FireORExplosionAffectmgOperabiltyof Safety Systems Required for Safe Shutdown I iHU4 Fire Within the Protected Area Boundary NOT Extinguished in 15 mm of Detection [ -3 j EAL reiIdalue EAL Threshold Value*
- 1. Fire or exploston causing damage to a Plant Vital Structure j1, Fire within or contiguous to aPlant Vital Structure (Table H- 1)1 (TableH-I ) or affecting one or more Safe Shutdown j AND Systems (Table 1-2) Fire is NOT extinguished in < 15 mm of EITHER, I AND .* Control Room notification Safe Shutdown System operability is required O
- Verfication ofalarm I : ~OR i l 2. Report by plant personnel of an unanticipated explosion within the Protected Area Boundary resulting in visible damage to a Plant Vital
__ i Structures (Table H-I) ubLeIl-ji Plant Vital Structures Table H-2: Safe Shutdown Systems Reactor Bldg Main #1 EDG Vault
- Isolation Condenier
- 4160 Safeguard Busses
- Control Room Ventilation
- ESW Intake Structure
- Contatnment Spray a Condensate Transfor
- Service Water Page 13 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS Many EALs are based on actual or potential degradation of fission product barriers because of the increased potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of increased radiological effluents or area radiation levels are appropriate symptoms for emergency classification.
At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions.
Increased area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.
Radiological Effluents Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits.
Projected offsite doses (based on effluent monitor readings) or actual offsite field measurements indicate doses or dose rates above classifiable limits.
Abnormal Rad Level Sustained general area radiation levels in excess of those indicating loss of control of radioactive materials or hindering access to vital plant areas also warrant emergency classification.
Coolant Activity Elevated activity in this process stream may be indicative of fuel clad degradation and is considered to be a precursor of more serious problems.
Elevated coolant activity in excess of Technical Specification limits may also be indicative of fuel clad degradation and is considered to be a precursor of more serious problems.
Irradiated Fuel Accidents This subcategory includes events that are indicative of uncontrolled level decrease or uncovery of spent fuel in either the Spent Fuel Pool or Reactor Cavity. This subcategory also addresses incidents associated with ISFSI.
Page 15 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RG1
'INITIATING CONDITION, Actual or Projected Site Boundary Dose Using Actual Meteorology:
> 5000 mRem CDE Child Thyroid EAL THRESHOLD VALUES
- 1. Radiological release in excess of Table R1 "General Emergency" threshold AND Release CANNOT be determined in < 15 min. (from time Table Rl threshold was exceeded) to be below Table R2 "General Emergency" thresholds OR
- 2. Radiological releases exceed ANY Table R2 "General Emergency" threshold.
Table RI: Effluent Monitor Thresholds Release Point Effluent Monitor General Emergency Torus/Drywell 2" Vent via SBGT to Main Stack Main Stack RAGEMS HRM >21.0 iCi/cc Reactor Building via Main Stack RAGEMS HRM > 3.0 liCi/cc SBGT to Main Stack Turbine Building via EF1-4 & EF1-33 to TB TB Stack RAGEMS HRM > 3.0 tCi/cc Stack HRM= High Range Monitor Page 16 of 122 Revision 0c
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS I EFFLUENTS RG1 (con't)
,EAL THRESHOLD VALUES (con't)
Table R2: Dose Assessment Thresholds Method General Emergency Sample Analysis N/A Field Team > 1000 mRem/hr Whole Body Monitoring* OR
> 5000 mRem CDE Child Thyroid Dose Assessment* > 1000 mRemO TEDE OR
> 5000 mRem CDE Child Thyroid
- At or beyond Site Boundary based on a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> release duration 1MODEk APtPLICABILITY .
ALL
'BASIS' Site Boundary - As specified in the ODCM.
Total Effective Dose Equivalent (TEDE) The sum of the deep dose equivalent (for external exposure) and the committed effective dose equivalent (for internal exposure) and 4 days of deposition exposure.
Committed Dose Equivalent (CDE) The Dose Equivalent to organs or tissues of reference that will be received from an intake of radioactive material by an individual during the 50-year period following the intake. The 5000 mRem integrated child thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for TEDE and child Thyroid Committed Dose Equivalent (CDE). Actual meteorology is used, since it gives the most accurate dose projection.
Table R1:
Effluent Monitors - Classification is based on the instantaneous release rate value if NO dose projections can be performed or verified within 15 minutes of meeting or exceeding the specified Release Rate value. Monitor indications (rounded) are calculated using the computerized dose model RAC with LOCA source terms applicable to each monitored pathway in conjunction with annual average (low end) meteorology and a one-hour release duration for Torus/DW 2" venting via SBGT or Turbine Building Vent and seven-hour release for Reactor Building venting via SBGT. The inputs are as follows:
Page 17 of 122 Revision Oc 0
F, Yplnn Nueloar Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RG1 (con't)
BASIS (con't)
Torus/DW Vent Reactor Building Turbine Building Main Stack Vent Main Stack Vent TB Stack Stability Class D D D Wind Speed 13.0 mph 13.0 mph 6.0 mph Wind Direction (from) 3160 (SE) 3160 (SE) 3160 (SE)
Accident LOCA LOCA LOCA Release Rate 20.9 pCi/cc 2.99 pCi/cc 3.08 pCi/cc Release Duration 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Time after S/D 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Table R2:
Field Team Monitoring - The values are for surveys or iodine air samples taken at or beyond the SITE BOUNDARY and are the most accurate indicator of the condition.
Field data are independent of release elevation and meteorology. The assumed release duration is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Direct reading iodine monitors are not available. Sampling of radioiodine by adsorption on charcoal media followed by field analysis are used for determining the iodine value.
REFERENCE(S)
- 1. 6630-ADM-4010.03 Oyster Creek Emergency Dose Calculation Manual
- 2. "Review of Calculations to Support EALs RGI and RS1" dated 11/22/02 to P.
Thompson from G. Seals, Radiological Engineer
- 3. "Maximum EAL Levels Effluent Radiation Monitors" dated 12/22/95 to T. Blount from J. Stevens - Engineer, Engineering & Design NUMARC IC' AGI DIFFERENCES
- 1. NUMARC IC AG1.2 - OCNS does not have telemetered perimeter monitors, therefore this EAL is not required Page 18 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RS1
'INITIATING CONDITION Actual or Projected Site Boundary Dose Using Actual Meteorology:
> 500 mRem CDE Child Thyroid
- EAL THRESHOLD VALUES
- 1. Radiological release in excess of Table R1 "Site Area Emergency" threshold AND Release CANNOT be determined in < 15 min. (from time Table RI threshold was exceeded) to be below Table R2 "Site Area Emergency" thresholds OR
- 2. Radiological releases exceed ANY Table R2 "Site Area Emergency" threshold Table RI: Effluent Monitor Thresholds Release Point Effluent Monitor Site Area Emergency Torus/Drywell vial SBGT toMan 2" Vent Stack Main Stack RAGEMS HRM > 2.1 gCi/cc via SBGT to Main Stack Reactor Building via Main Stack RAGEMS HRM >0. 3 gCi/cc SBGT to Main Stack Turbine Building via TB Stack RAGEMS HRM > 0.3 gCi/cc EF1-4 & EF1-33 to TB Stack HRM= High Range Monitor Page 19 of 122 Revision 0c
Avfji. Crdr 1%TipIiiir ffmnn Anniw fA .L efoir Annav Vý ý I U -IV "ý I ý"A Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RS1 (con't)
'EAL THRESHOLD VALUES (con't)
Table R2: Dose Assessment Thresholds Method Site Area Emergency Sample Analysis N/A Field Team > 100 mRem/hr Whole Body Monitoring* OR
> 500 mRem CDE Child Thyroid Dose Assessment* O100RmRem TEDE OR
> 500 mRem CDE Child Thyroid
- At or beyond Site Boundary based on a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> release duration MODE APPLICwA.B..L..ITY.
ALL 1BASIS Site Boundary - As specified in the ODCM.
Total Effective Dose Equivalent (TEDE) The sum of the deep dose equivalent (for external exposure) and the committed effective dose equivalent (for internal exposure) and 4 days of deposition exposure.
Committed Dose Equivalent (CDE) The Dose Equivalent to organs or tissues of reference that will be received from an intake of radioactive material by an individual during the 50-year period following the intake. The 500 mRem integrated child thyroid dose was established in consideration of the 1:5 ratio of the EPA Protective Action Guidelines for TEDE and child Thyroid Committed Dose Equivalent (CDE). Actual meteorology is used, since it gives the most accurate dose projection.
Table Rl:
Effluent Monitors - Classification is based on the instantaneous release rate value if NO dose projections can be performed or verified within 15 minutes of meeting or exceeding the specified Release Rate value. Monitor indications (rounded) are calculated using the computerized dose model RAC with LOCA source terms applicable to each monitored pathway in conjunction with annual average (low end) meteorology and a one-hour release duration for Torus/DW 2" venting via SBGT or Turbine Building Vent and seven-hour release for Reactor Building venting via SBGT. The inputs are as follows:
Page 20 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RS1 (con't)
BASIS (con't)
Torus/DW Vent Reactor Building Turbine Building Vent Main Stack Vent Main Stack TB Stack Stability Class D D D Wind Speed 13.0 mph 13.0 mph 6.0 mph Wind Direction (from) 3160 (SE) 3160 (SE) 3160 (SE)
Accident LOCA LOCA LOCA Release Rate 2.09 ptCi/cc 0.299 ptCi/cc 0.308 jiCi/cc Release Duration I hour 7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Time after S/D 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Table R2:
Field Team Monitoring - The values are for surveys or iodine air samples taken at or beyond the SITE BOUNDARY and are the most accurate indicator of the condition.
Field data are independent of release elevation and meteorology. The assumed release duration is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Direct reading iodine monitors are not available. Sampling of radioiodine by adsorption on charcoal media followed by field analysis are used for determining the iodine value.
This event will be escalated to a General Emergency when actual or projected doses exceed EPA-400-R-92-001 Protective Action Guidelines per IC RG1.
- REFERENCE(S)
- 1. 6630-ADM-4010.03 Oyster Creek Emergency Dose Calculation Manual
- 2. "Review of Calculations to Support EALs RG1 and RSl" dated 11/22/02 to P.
Thompson from G. Seals, Radiological Engineer
- 3. "Maximum EAL Levels Effluent Radiation Monitors" dated 12/22/95 to T. Blount from J. Stevens - Engineer, Engineering & Design NUMARC IC AS1 DIFFERENCES
- 1. NUMARC IC AS1.2 - OCNS does not have telemetered perimeter monitors, therefore this EAL is not required.
Page 21 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RAI INITIATING CONDITION Release > 200 X ODCM Limit for > 15 min.
EAL THRESHOLD VALUES
- 1. Unplanned radiological release lasting > 15 min. in excess of Table R1 "Alert" thresholds AND Releases CANNOT be determined in < 15 min. (from time Table R1 threshold was exceeded) to be below Table R2 "Alert" thresholds OR
- 2. Unplanned radiological releases lasting > 15 minutes in excess of ANY Table R2 "Alert" threshold Table RI: Effluent Monitor Thresholds Release Point/Monitor Alert
- Turbine Bldg. Noble Gas RAGEMS Via EF1-4 & EFl-33 > 200 X Hi Hi alarm setpoint
- Service Water > 3.13 E4 cpm Table R2: Dose Assessment Thresholds Method Alert
> 200 X ODCM 4.6.1.1.4 Sample Analysis OR
> 200 X ODCM 4.6.1.1.5
> 10 mRem/hr Whole Body Field Team Monitoring* OR
> 34 mRem CDE Child Thyroid
> 10 mRem/hr TEDE Dose Assessment* OR
> 34 mRem CDE Child Thyroid
- At or beyond Site Boundary based on a I hour release duration Page 22 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RAI - Cont'd MODE APPLICABILITY ALL
'BASIS Unplanned - Not the result of an intended evolution and requiring corrective or mitigative actions Sustained - A sustained unplanned release of this greater magnitude that cannot be terminated in 15 minutes represents an uncontrolled situation that is an actual or potential substantial degradation of the level of safety of the plant. The degradation in plant control implied by the fact that the release cannot terminated in 15 minutes is the primary concern. The Emergency Director should not wait until 15 minutes has elapsed, but should declare an event as soon as the release is determined to be uncontrolled or projected to be non-isolable within 15 minutes.
Table Rl:
Effluent Monitors - Monitor indications are calculated based on the methodology of the site Offsite Dose Calculation Manual (ODCM). The Hi Hi alarm set point for the Turbine Building gaseous effluent monitor is set conservatively to indicate when a potential release may approach ODCM limits assuming multiple release points. Use of multiples of this conservative set point in establishing a monitor reading will not cause an inappropriate event classification since this EAL requires the magnitude of the monitor reading to be two hundred times the set point, sustained for >15 minutes, and assessment by a dose projection indicating an offsite dose rate in excess of two hundred times ODCM limits. The Main Stack value (CPS A) is calculated on a periodic basis consistent with the multiple of 200 times ODCM limits. Service Water effluent monitor value of 3.13 E4 cpm is consistent with 200 times ODCM allowed concentrations.
Table R2:
It is intended that the event be declared as soon as it is determined that the release will exceed two hundred times ODCM for greater than 15 minutes.
Samples Analysis - The calculation called for in this EAL should also be conducted whenever a liquid release occurs for which a radioactive discharge permit wasn't prepared or that exceeds the conditions on the permit (e.g. minimum dilution, alarm set points, etc).
Field Team Monitoring - The values are for surveys or iodine air samples taken at or beyond the SITE BOUNDARY and are the most accurate indicator of the condition.
Field data are independent of release elevation and meteorology. The assumed release duration is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Direct reading iodine monitors are not available. Sampling of radioiodine by adsorption on charcoal media followed by field analysis are used for determining the iodine value.
Page 23 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RA1 - Cont'd BASIS - Cont'd Dose Assessment - This EAL includes a 15 minute average for the dose projection with the release point radiation monitor above two hundred times the Hi Hi alarm set point value for the entire 15 minutes. It is not intended that the release be averaged over 15 minutes, but exceed threshold for 15 minutes.
The TEDE is calculated by dividing the yearly allowable ODCM limit (500 mRem/yr.) by the number of hours per year (8760 hr/yr.), and then multiplying by a factor of 200 times ODCM TEDE = 200x(Tech Spec Limit)/(hours per year)
= 200(500 mRem/yr.)/(8760 hr/yr.)
= 11.4 mRem/hr (rounded to 10 mR/hr)
The CDE Thyroid is calculated by dividing the yearly allowable ODCM limit (1500 mRem/yr.) by the number of hours per year (8760 hr/yr.), and then multiplying by a factor of 200 times ODCM.
CDE = 200x(Tech Spec Limit)/(hours per year)
= 200(1500 mRem/yr.)/(8760 hr/yr.)
= 34.2 mRem/hr (rounded to 34 mRlhr)
Releases in excess of 11 mRem/hr TEDE, 34 mRem CDE Thyroid, or 200 times the ODCM limits that continue for > 15 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The primary concern is the final integrated dose [100 times greater than the Unusual Event] and the degradation in plant control implied by the fact that the release was not isolated within 15 minutes.
This event will be escalated to a Site Area Emergency when actual or projected doses are determined to exceed 10CFR20 annual average population exposure limits per IC RS 1.
REFERENCE(S)
- 1. 2000-ADM-4532.04 Oyster Creek Offsite Dose Calculation Manual
- 2. "Maximum EAL Levels Effluent Radiation Monitors" dated 12/22/95 to T. Blount from J. Stevens - Engineer, Engineering & Design
- 1. NUMARC IC AA1.3 - OCNS does not have telemetered perimeter monitors, therefore this EAL is not required.
- 2. NUMARC IC AA1.4 - OCNS does not use automatic initiation of real time dose assessment therefore this EAL is not required.
Page 24 of 122 Revision 0c
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RU1
.INITIATING CONDITION Release > 2 X ODCM Limit for > 60 min.
EAL THRESHOLD VALUE'
- 1. Unplanned radiological release lasting > 60 min. in excess of Table RI "Unusual Event" threshold AND Releases CANNOT be determined in < 60 min. (from time Table RI threshold was exceeded) to be below Table R2 "Unusual Event" thresholds OR
- 2. Unplanned radiological releases lasting > 60 min. in excess of ANY Table R2 "Unusual Event" threshold Table RI: Effluent Monitor Thresholds Release Point/Monitor Unusual Event
"* Main Stack Noble Gas RAGEMS Torus/Drywell 2" Vent via SBGT > CPS U Reactor Bldg. via SBGT > CPS U
"* Turbine Bldg. Noble Gas RAGEMS Via EF1-4 & EFl-33 > 2 X Hi Hi alarm setpoint
- Service Water > 4.15 E2 cpm Table R2: Dose Assessment Thresholds Method Unusual Event
> 2 X ODCM 4.6.1.1.4 Sample Analysis OR
> 2 X ODCM 4.6.1.1.5 Field Team Monitoring* N/A
> 0.10 mR/hr TEDE Dose Assessment* OR
> 0.34 mRem CDE Child Thyroid
- At or beyond Site Boundary based on a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> release duration Page 25 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RUl - Cont'd
'MODE APPLICABILITY ALL BASIS--"'
Table RI:
Effluent Monitors - Monitor indications are calculated based on the methodology of the site Offsite Dose Calculation Manual (ODCM). The Hi Hi alarm set point for the Turbine Building gaseous effluent monitor is set conservatively to indicate when a potential release may approach ODCM limits assuming multiple release points. Use of multiples of this conservative set point in establishing a monitor reading will not cause an inappropriate event classification since this EAL requires the magnitude of the monitor reading to be two times the set point, sustained for >60 minutes, and assessment by a dose projection indicating an offsite dose rate in excess of two times ODCM limits. The Main Stack value (CPS U) is calculated on a periodic basis consistent with the multiple of 2 times ODCM limits. Service Water effluent monitor value of 4.15 E2 cpm is consistent with 2 times ODCM allowed concentrations.
Table R2:
It is intended that the event be declared as soon as it is determined that the release will exceed two times ODCM for greater than 60 minutes.
Sample Analysis - The calculation called for in this EAL should also be conducted whenever a liquid release occurs for which a radioactive discharge permit wasn't prepared or that exceeds the conditions on the permit (e.g. minimum dilution, alarm set points, etc).
Dose Assessment - This EAL includes a 60 minute average for the dose projection with the release point radiation monitor above two times the Hi Hi alarm set point value for the entire 60 minutes. It is not intended that the release be averaged over 60 minutes, but exceed threshold for 60 minutes.
The TEDE is calculated by dividing the yearly allowable ODCM limit (500 mRem/yr.) by the number of hours per year (8760 hr/yr.), and then multiplying by a factor of 2 times the ODCM limit.
TEDE = 2x (Tech Spec Limit)/(hours per year)
= 2 (500 mRem/yr.)/(8760 hr/yr.)
= 0.114 mRem/hr (rounded to 0.10 mR/hr)
The CDE is calculated by dividing the yearly allowable ODCM limit (1500 mRem/yr.) by the number of hours per year (8760 hr/yr.), and then multiplying by a factor of 2 times ODCM limit.
CDE = 2x (Tech Spec Limit)/(hours per year)
= 2 (1500 mRem/yr.)/(8760 hr/yr.)
= 0.342 mRem/hr (rounded to 0.34 mR/hr)
Page 26 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RU1 - Cont'd
'BASIS - Cont'd Releases in excess of 0.10 mRem/hr TEDE, 0.34 mRem/hr CDE Child Thyroid, or 2 x ODCM limits that continue for > 60 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The final integrated dose is very low and is not the primary concern. Rather it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes.
REFEENCE(S)
- 1. 2000-ADM-4532.04 Oyster Creek Offsite Dose Calculation Manual
- 2. "Maximum EAL Levels Effluent Radiation Monitors" dated 12/22/95 to T. Blount from J. Stevens - Engineer, Engineering & Design
- 1. NUMARC IC AU1.3 - OCNS does not have telemetered perimeter monitors, therefore this EAL is not required.
- 2. NUMARC IC AU1.4 - OCNS does not use automatic initiation of real time dose assessment therefore this EAL is not required.
Page 27 of 122 Revision 0c
n .L efav Pvag-L- V"Aaar Qfat4tin A ininwv lwv-d-lnin N"Aaaw Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS I EFFLUENTS RA2 INITIATING CONDITION In-Plant Radiation Levels Impede Plant Operations EAL THRESHOLD VALUES
- 1. Radiation readings > 15 mR/hr in EITHER of the following:
"* Main Control Room OR
"* Central Alarm Station OR
- 2. In-plant radiation readings > 1 R/hr in areas requiring access in order to maintain safe operation or perform a safe shutdown MODE APPLICABILITY ALL BASIS This EAL addresses elevated radiation levels that impede necessary access to operator stations, or other areas containing equipment that must be operated manually in order to maintain safe operation or to perform a safe shutdown. The concern of the EAL is a loss of control of radioactive material causing high radiation levels.
The impaired ability to operate the plant is to be considered as the actual or potential substantial degradation of the level of safety of the plant. The cause of the rise in radiation levels is not the major concern of this EAL. For example, a dose rate of 15 mR/hr in the control room or high radiation monitor readings may also be indicative of high dose rates in the containment due to a LOCA. In this latter case, the fission product barrier table may indicate a SAE or GE.
Threshold Value 1 - The value of 15 mRem/hr is derived from the general design criteria (GDC) value of 5 REM in 30 days with adjustment for expected occupancy times.
Although Section III.D.3 of NUREG-0737 "Clarification of TMI Action Plan Requirements" provides that the 15 mRem/hr value can be averaged over the 30 days, the value is used here without averaging, as a 30 day duration implies an event potentially more significant than an ALERT.
Plant normal and emergency procedures may be implemented without requiring any areas except the Control Room and Central Alarm Station to be continuously occupied.
Page 28 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RA2 - Cont'd BASIS - Cont'd Threshold Value 2 - Areas requiring infrequent access and dose rate values are based on those specified in the Emergency Operating Procedure (EMG-3200.1 1) Secondary Containment Control. The single value of I R/hr (Maximum Safe Operating Value) was selected because it is based on radiation levels which result in exposure control measures intended to maintain doses within normal occupational exposure guidelines and limits (i.e.,
10 CFR 20), and in doing so, will impede necessary access. Stay times for levels up to that value are, generally several minutes, enough time to enter an area and manually operate the equipment. Dose rates > I R/hr may impede necessary access.
REFERENCE(S).
- 1. EMG-3200.11 Secondary Containment Control NUMARC IC ...
AA3 DIFFERENCES None Page 29 of 122 Revision 0c
Ovtr (rppk NuuiInr tfltinn Annpr "lq":,tpl n n ]*Tllrlp* r Ovster Creek Nuclear Station Annex VxPinn N"Apar Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS I EFFLUENTS RU2 INITIATING CONDITION Rise in Plant Radiation Levels By a Factor of 1000 IELA'THRESHOLD' VALUE Valid area radiation monitor readings indicate an unplanned rise by a factor of 1000 over normal levels as detected by either permanent or temporarily installed radiation monitors or by manual survey MODE APPLICABILITY ALL BASIS Unplanned - Not the result of an intended evolution and requiring corrective or mitigative actions.
Normal Levels - Normal radiation levels can be considered as the highest reading in the past 24-hour period, excluding the current peak value.
Valid - An area monitor is considered to be valid when it is verified by:
"* An instrument channel check indicating that the monitor has not failed;
"* Indications on related or redundant instrumentation, or
"* Direct measurement by plant personnel.
Classification of an UNUSUAL EVENT is warranted as a precursor to more serious events. The concern of this EAL is the loss of control of radioactive material representing a potential degradation of the level of safety of the plant. The Threshold Value tends to have a long lead-time relative to a radiological release and thus the threat to public health and safety is very low.
Area radiation measurements should include both permanent and temporarily installed area radiation monitoring instruments as well as manual area radiation surveys but excludes Containment High Range Radiation Monitors (CHRRMS). Increases in area radiation in normally high radiation areas would be classified under EAL RA2.
REFERENCE(S)
None
,NUMARC IC AU2 DIFFERENCES None Page 30 of 122 Revision 0e
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RU3 INITIATING CONDITION High Off-gas Radiation Levels
,EAL THRESHOLD VALUES Off-gas radiation > Hi-Hi alarm value for > 15 min.
MODE APPLICABILITY:.
1,2 IBASIS:
The steam jet air ejector discharge (off-gas) radiation monitor RN05E(F) in the Control Room would be one of the first indicators of a potentially degrading core. The high-high alarm is nominally set below the Technical Specification limit. The Hi Hi alarm results in the closure of V-7-3 1, V-7-29 and OG-AOV-001A (001 B) to isolate the off-gas system at the stack and trip the mechanical vacuum pump (if running) after a 15 minute time delay.
This instrument takes a sample before the recombiner. This indicator of elevated activity is considered to be a precursor of more serious problems. The Technical Specification limit reflects a degrading or degraded core condition.
If operator action to reduce off-gas radiation levels is not successful within the 15 minute time frame, this level of steam line activity is assumed to be indicative of the release of gap activity to the coolant. The mechanics that caused off-gas radiation to rise to this level indicate there are a degradation of Fuel Clad integrity and thus a threat to the Fuel Clad Barrier.
This EAL is NOT intended to apply to cases caused by resin intrusion or other known factors.
REFERENCE(S)
- 1. 2000-ABN-3200.26, Increase in Off Gas Activity
- 1. The MODE applicability [1,2] is a deviation from NUMARC [all] in that, the SJAE Radiation Monitor will only be a valid indication of Fuel Clad Degradation in MODE's [1, 2].
Page 31 of 122 Revision Oc
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RU4 INITIATING CONDITION High Coolant Activity EAL THRESHOLD VALUES Reactor Coolant activity > 0.2 uCi/gm DEI
'MOE APPLICABiLITY.:
ALL BASIS: (References)
Coolant activity in excess of Technical Specifications (> 0.2uCi/gm) is considered to be a precursor of more serious problems. The Technical Specification limit reflects a degrading or degraded core condition. This level is chosen to be above any possible short duration spikes under normal conditions. An Unusual Event is only warranted when actual fuel clad damage is the cause of the elevated coolant sample (as determined by laboratory confirmation). However, fuel clad damage should be assumed to be the cause of elevated Reactor Coolant activity unless another cause is known, e.g., Reactor Coolant System chemical decontamination evolution (during shutdown) is ongoing with resulting high activity levels.
This event will be escalated to an Alert when Reactor Coolant activity exceeds 300 jtCi/gm Dose Equivalent Iodine 131 per Fission Product Barrier Matrix FC. 1.
tREFERENCE(S). .
- DIFFERENCES None Page 32 of 122 Revision Oc
(vcf.r Criu'Jr N,,r1no, tifnn ArunDv
" .L efav- rv-g%,mL- WizAaaw Qfa+in" A ""ný V -1 I -- 1V -1 I ý"l Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RA5
,INITIATING CONDITION.
Major Damage OR Uncovering of Spent Fuel EAL THRESHOLD VALUES
- 1. Valid unanticipated Hi alarm on one or more Refuel Floor ARMs (Table R-3)
- 2. Report of visual observation that irradiated fuel in the Spent Fuel Pool or Fuel Transfer Canal is uncovered Table R-3 Refuel Floor ARMs
- C-5, Crit Mon
- C-10, North Wall
- C-9, North Wall
- B-9, Open Floor
,MODE APPLICABILITY ALL 1BASIS: (References)
Offsite doses during these accidents would be well below the EPA Protective Action Guidelines and the classification as an Alert is therefore appropriate. This radiation level could also be caused by an inadvertent criticality and is included even though the probability of this event occurring is low. Radiation levels that rise above the Hi alarm set point, which were expected during a planned evolution, should not cause an Alert to be declared.
This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage (ISFSI), which is addressed in RU7.
NUREG-0818, "Emergency Action Levels for Light Water Reactors," forms the basis for this EAL. The areas where irradiated fuel is located forms the basis for Table R-3.
Unexpected radiation levels, significantly higher than the normal background will generally indicate a fuel handling accident or loss of water covering the irradiated fuel.
Readings may be from refuel floor Area Radiation Monitors or taken during a qualified radiological survey.
Page 33 of 122 Revision 0c
1U-nil,* W-0 la1-Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RA5 - Cont'd BASIS: (References)
- Cont'd There is time available to take corrective actions, and there is little potential for substantial fuel damage. In addition, NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82," July 1987, indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and that risk of injury is low.
REFERENCE(S)
- 1. 2000-RAP-3024.01 1OF NSSS Alarm Response Procedures
,NUMARC IC AA2 "DIFFERENCES None Page 34 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS I EFFLUENTS RU5 INITIATING CONDITION Potential Damage OR Uncovering of Spent Fuel EAL THRESHOLD VALUE Uncontrolled water level drop in the Spent Fuel Pool with ALL irradiated fuel assemblies remaining covered by water
'MODE APPLICABILITY ALL BASIS: (References)
Uncontrolled - An unexplained level drop that cannot be quickly terminated and is not the result of a planned evolution.
This event tends to have a long lead time relative to potential for radiological release outside the site boundary, thus impact to public health and safety is very low.
In light of Reactor Cavity Seal failure incidents at two different PWRs and loss of water in the Spent Fuel Pit/Fuel Transfer Canal at a BWR all occurring since 1984, explicit coverage of these types of events via this EAL is appropriate given their potential for elevated doses to plant staff. Classification as an Unusual Event is warranted as a precursor to a more serious event.
This event will be escalated to an Alert as a result of uncovery of a fuel assembly and/or indication of high radiation levels on the refueling floor.
REFERENCE(S)
None
'NUMARC IC AU2 DIFFERENCES None Page 35 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RA6 INITIATING CONDITION Loss of Water Level That Has OR Will Uncover Irradiated Fuel in the Reactor Cavity EAL THRESHOLD VALUES Report of visual observation that irradiated fuel in the Reactor Cavity is or will be uncovered.
MODE APPLICABILITY 4
BASIS: (References)
This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage, which is discussed in RU7.
NUREG-0818, "Emergency Action Levels for Light Water Reactors," forms the basis for this EAL.
There is time available to take corrective actions, and there is little potential for substantial fuel damage. In addition, NUREG/CR-4982, "Severe Accident in Spent Fuel Pools in Support of Generic Safety Issue 82," July 1987, indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and that risk of injury is low.
REFERENCE(S)
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RU6 INITIATING CONDITION Uncontrolled Water Level Drop in Reactor Cavity
- EAL THRESHOLD VALUE Unexpected Skimmer Surge Tank Lo-Lo level alarm AND Visual observation of an uncontrolled drop in water level below the fuel pool skimmer surge tank inlet
'MODE APPLICABILITY 4
BASIS (References)
Unexpected - An alarm that is not a result of a planned evolution Uncontrolled - An unexplained level drop that cannot be quickly terminated and is not the result of a planned evolution A drop in the Spent Fuel Pool level or the RPV [when in refueling and flooded up with the gates removed] will result in a control room annunciator Skimmer Surge Tank Level Lo-Lo Alarm. This alarm is validated with visual observation of a lowering Spent Fuel Pool level.
If the spent fuel pool level drops below the inlet to the skimmer surge tank, without a planned event such as removing a large piece of equipment, there must be a leak in the spent fuel pool or the RPV. This event has a long lead time relative to potential for radiological release outside the site boundary, thus the impact to public health and safety is very low. Classification as an Unusual Event is warranted as a precursor to a more serious event.
In light of Reactor Cavity Seal failure incidents at two different PWRs and loss of water in the Spent Fuel Pit/Fuel Transfer Canal at a BWR all occurring since 1984, explicit coverage of these types of events via this EAL is appropriate given their potential for elevated doses to plant staff. Classification as an Unusual Event is warranted as a precursor to a more serious event.
This event will be escalated to an Alert as a result of uncovery of a fuel assembly and/or indication of high radiation levels on the refueling floor.
Page 37 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS I EFFLUENTS RU6 - Cont'd REFERENCE(S)
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS I EFFLUENTS RU7 INITIATING CONDITION Independent Spent Fuel Storage Installation (ISFSI)
EAL THRESHOLD VALUE Radiation readings > 10 times normal at ANY of the following ISFSI locations:
"* on contact with roof OR
"* on contact with shield door centerline OR
"* on contact with shield wall iMODE APPLICABILITY ALL
'BASIS: (References)
This EAL applies to potential emergency conditions, which might develop during use of the Independent Spent Fuel Storage Installation and dry cask storage system. This EAL provides for an Unusual Event classification, which may be entered in the event that conditions occur which have the potential for damaging or degrading the fuel, but no releases of radioactive material requiring offsite response or monitoring are expected.
Consistent with the NUMARC guidance, escalations above the Unusual Event are not warranted.
Accidents associated with the dry cask storage system include natural and man-made events that are postulated to affect the storage system. The limiting impacts to the system include loss of shielding capability and loss of confinement. The loss of shielding results in higher direct radiation to the environment from the cask while the loss of confinement results in a release of materials from within the cask to the environment at a postulated leak rate.
The threshold of 10 times normal is significantly above normal operating values and is high enough to eliminate false indications. It is indicative of some failure or external event resulting in a dose rate problem requiring response actions.
Loss of confinement for the dry storage system is evaluated in the Safety Analysis Report. Scenarios are considered for both off-normal conditions and for hypothetical accident conditions. In the extremely unlikely event that one of these scenarios did occur, the event would be addressed by other EALs under Category R, "Abnormal Radiological Levels / Effluents."
Page 39 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RU7 - Cont'd REFERENCE(S)
- 1. OCNS ISFSI Certificate of Compliance
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER DEGRADATION Fission Product Barrier EALs EALs defined in this category represent threats to a defense in depth design that precludes the release of highly radioactive fission products to the environment. The design relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.
The primary fission product barriers are:
"* Fuel Clad (FC): The zirconium tubes which house the ceramic uranium oxide pellets along with the end plugs which are welded into each end of the fuel rods comprise the fuel cladding.
" Reactor Coolant System (RCS): The reactor vessel shell, vessel head, vessel nozzles and penetrations and all primary systems directly connected to the reactor vessel up to the first containment isolation valve comprise the RCS.
" Primary Containment (PC): The drywell, torus, the interconnections between the two, and all isolation valves required to maintain primary containment integrity under accident conditions comprise the containment barrier.
Although the secondary containment (reactor building) serves as an effective fission product barrier by minimizing ground level releases, it is not considered as a fission product barrier for the purpose of emergency classification.
The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1.
Although the logic used for these initiating conditions appears overly complex, it is necessary to reflect the following considerations:
"* The Fuel Clad barrier and the RCS barrier are weighted more heavily than the Primary Containment barrier. Unusual Event ICs associated with RCS and Fuel Clad barriers are addressed under the other plant condition EALs.
" At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are from General Emergency. For example, if the Fuel Clad barrier and RCS barrier "Loss" EALs existed, this would indicate to the Emergency Director that, in addition to offsite dose assessments, the ED must focus on continual assessments of radioactive inventory and containment integrity. If, on the other hand, both Fuel Clad barrier and RCS barrier "Potential Loss" EALs existed, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.
"* The ability to escalate to higher emergency classes as an event gets worse must be maintained. For example, RCS leakage steadily rising would represent an increasing risk to public health and safety.
Page 41 of 122 Revision Oc
Ovq*ter Creek Nuclear S9tation Annexy Exelon Nnclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER DEGRADATION Fission Product Barrier ICs must be capable of addressing event dynamics. An IMMINENT (i.e., within 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) Loss or Potential Loss should result in a classification as if the affected threshold(s) are already exceeded, particularly for the higher emergency classes.
If a "Loss" condition is satisfied, the "Potential Loss" category can be considered satisfied.
This is also applicable to conditions where there is a "Loss" indication with no corresponding "Potential Loss" condition.
For all conditions listed in Fission Product Barrier Table, the barrier failure column is only satisfied if it fails when called upon to mitigate an accident. For example, failure of both containment isolation valves to isolate with a downstream pathway to the environment is only a concern during an accident. If this condition exists during normal power operations, it will be an active Technical Specification Action Statement. However, during accident conditions, this will represent a breach of Primary Containment.
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,pynilin N11.4l~t Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER DEGRADATION Table F-1 Fission Product Barrier Matrix
- 1. FUEL CLAD BARRIER LOSS POTENTIAL LOSS
- a. RPV Water Level
- b. Drvwell Radiation Monitonng
- 1. Containment Hi-Range Radiation Monitoring System (CHRRMS) > 440 R/hr NA
- c. Drywell Pressure NA NA
- d. Breached / Bypassed (Primary Coolant Activity Level)
- 1. Coolant activity > 300 jiCilgm (DEI)
NA
- e. Containment Hydrogen Concentration NA NA f Emergency Director Judgment
- 1. ANY condition in the judgment of the Emergency Director that indicates Loss or Potential Loss of the Fuel Clad barrier Page 43 of 122 Revision 0e
Ovtr Crk NiWlP!Ir ttmnn Annv .VOlnin I~ld~nal
.L stpr rrpek N"elpar Qfnflniri Alnlnpv Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER DEGRADATION
OR NA CANNOT be determined b Drvwell Radiation Monitoring
- 1. Containment Ili-Range Radiation Monitoring System (CttRRMS) >45 R/hr NA
- c. DrUwell Pressure
- 1. Drywell pressure > 3.0 psig AND NA Indication of a RCS leak inside drywell d, Breached I Bypassed
- 1. Unisolable Main Steam Line break outside containment 3. RCS leakage >50 gpm OR OR
- 2. Unisolable Isolation Condenser tube rupture 4. Unisolable primary system leakage outside of drywell as indicated by exceeding EITHER of the following in one or more areas requiring a scram:
EMG-3200.11 Max Normal Temperature OR EMG-3200.11 Max Normal Radiation Level e Containment Ilydrogen Concentration NA NA f.Emergency Director Judgment
- 1. ANY condition in the judgment of the Emergency Director that indicates Loss or Potential Loss of the RCS barrier Page 44 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelnn Niiel'sr Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER DEGRADATION
- 3. PRIMARY CONTAINMENT BARRIER LOSS POTENTIAL LOSS
- 1. Containment Hi-Range Radiation Monitoring System (CIIRRMS) > 2.0E+4 NA R/hr c Drywell Pressure
- 1. Rapid, unexplained drop in drywell pressure following an initial rise OR 3, Drywell pressure > 44 psig
- 2. Drywell pressure response not consistent with LOCA conditions indicating a containment breach
- d. Breached/Bypassed
- 1. Failure of all isolation valves in ANY one line penetrating Primary Containment to close when required AND NA Downstream pathway exists to environment
- 2. Intentional venting per EMG-3200 02 is required with Drywell pressure > 3.0 psig NA
- 3. Unisolable primary system leakage outside of drywell as indicated by exceeding EITHER of the following in one or more areas requiring a scram:
EMG-3200.11 Max Normal Temperature NA OR EMG-3200.11 Max Normal Radiation Level e Containment Hydrogen Concentration
- 1. Containment 112 concentration > 6%
NA AND Containment 02 concentration > 5%
f Emergency Director Judgment
- 1. ANY condition in the judgment of the Emergency Director that indicates Loss or Potential Loss of the Primary Containment barrier Page 45 of 122 Revision Oc
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER FG1 INITIATING CONDITION Loss of 2 Fission Product Barriers and Loss or Potential Loss of the Third Barrier EAL THRESHOLD VALUE Comparison of conditions / values with those listed in Fission Product Barrier Matrix, Table F-i, indicates:
LOSS of ANY two barriers AND LOSS or POTENTIAL LOSS of a third barrier MODE APPLICABILITY 1 and 2 BASIS Conditions / events required to cause the loss of 2 Fission Product Barriers with the potential loss of the third could reasonably be expected to cause a release beyond the immediate site area exceeding EPA Protective Action Guidelines.
Guidance for development of this EAL was taken from Recognition Category F in the NUMARC/NESP-007 Methodology for Development of Emergency Action Levels.
A barrier LOSS shall also constitute a POTENTIAL LOSS for classification purposes.
Refer to Table F-1 for Fuel Clad, RCS and Primary Containment loss and potential loss indicators and bases.
REFERENCE(S)
- 1. NUMARC/NESP-007, Revision 2, Section 3.4 & Table 5-F-i NUMARC IC FGI - Table 5-F-i Site Area Emergency
ýDIFFERENCES None Page 46 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER FS1 INITIATING CONDITION Loss or Potential Loss of BOTH Fuel Clad AND RCS Barriers OR Loss or Potential Loss of EITHER the Fuel Clad OR RCS Barrier, AND a Loss of Another Barrier EAL THRESHOLD VALUE Comparison of conditions / values with those listed in Fission Product Barrier Matrix, Table F-I, indicates:
LOSS or POTENTIAL LOSS of BOTH Fuel Clad AND RCS Barriers OR LOSS or POTENTIAL LOSS of EITHER the Fuel Clad OR RCS Barrier AND a LOSS of Primary Containment Barrier
'MODE'APPLICABILITY.
1 and 2
'BASIS Loss of 2 Fission Product Barriers would be a major failure of plant systems needed for protection of the public.
Guidance for development of this EAL was taken from Recognition Category F in the NUMARC/NESP-007 Methodology for Development of Emergency Action Levels.
A barrier LOSS shall also constitute a POTENTIAL LOSS for classification purposes.
Refer to Table F-1 for Fuel Clad, RCS and Primary Containment loss and potential loss indicators and bases.
REFERENCE(S).
- 1. NUMARC/NESP-007, Revision 2, Section 3.4 & Table 5-F-1 NUMARC IC FS1 - Table 5-F-1 Site Area Emergency DIFFERENCES None Page 47 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER FA1 INITIATING CONDITION Loss or Potential Loss of EITHER Fuel Clad OR RCS Barriers EAL THRESHOLD VAiLUE.
Comparison of conditions / values with those listed in Fission Product Barrier Matrix, Table F-i, indicates:
ANY LOSS or POTENTIAL LOSS of Fuel Clad barrier OR ANY LOSS or POTENTIAL LOSS of Reactor Coolant System barrier MODE APPLICABILTY 1 and 2 IBASIS The Fuel Cladding and the Reactor Coolant System are weighted more heavily than the Containment Barrier.
A LOSS or POTENTIAL LOSS of either the Fuel Cladding or the Reactor Coolant System would be a substantial degradation in the level of plant safety.
Guidance for development of this EAL was taken from Recognition Category F in the NUMARC/NESP-007 Methodology for Development of Emergency Action Levels.
Refer to Table F-i for Fuel Clad, RCS and Primary Containment loss and potential loss indicators and bases.
REFERENCE(S)
- 1. NUMARC/NESP-007, Revision 2, Section 3.4 & Table 5-F-1 NUMARC iC FS I - Table 5-F-1 Alert DIFFERENCES None Page 48 of 122 Revision Oc
Exelan Nnclaar 0 ster Creek Nuclear Station Annexi~ Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER FUl INITIATING CONDITION ANY Loss or Potential Loss of Containment EAL THRESHOLD VALUE Comparison of conditions / values with those listed in Fission Product Barrier Matrix, Table F-i, indicates:
ANY LOSS or POTENTIAL LOSS of Primary Containment
ýMODE APPLICABILITY 1 and 2 iBASIS Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier.
Unlike the Fuel Clad and RCS (the loss of either of which results in an Alert) loss of containment in and of itself does not result in the relocation of radioactive materials or the potential for loss of core cooling capability. However, loss or potential loss of containment in combination with loss or potential loss of either Fuel Clad or RCS barriers results in declaration of a Site Area Emergency.
Guidance for development of this EAL was taken from Recognition Category F in the NUMARC/NESP-007 Methodology for Development of Emergency Action Levels.
Refer to Table F-1 for Fuel Clad, RCS and Primary Containment loss and potential loss indicators and bases.
REFERENCE(S)
- 1. NUMARC/NESP-007, Revision 2, Section 3.4 & Table 5-F-1
?NUMARCIC FS I - Table 5-F-I Unusual Event
-DIFFERENCES None Page 49 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER TABLE F-1 Barrier Threshold Bases 1.0 FUEL CLAD BARRIER BASES
- a. RPV Water Level Loss: 1. RPV level <-30" TAF OR CANNOT be determined The specified RPV water level is the Minimum Steam Cooling RPV Water Level (MSCRWL) and is used in EOPs to indicate challenge to core cooling. The MSCRWL is the lowest RPV water level at which the submerged portion of the reactor core will generate sufficient steam to prevent any clad in the uncovered portion of the core from heating to 1500'F; the threshold temperature of fuel clad perforation. This water level is utilized to preclude fuel damage when RPV water level is below the top of active fuel (TAF).
The MSCRWL appears in the RPV CONTROL - WITH ATWS procedure when RPV water level is intentionally lowered to reduce reactor power. When RPV water level is deliberately lowered, power instabilities may produce noticeable oscillations in RPV water level and make it difficult to maintain water level exactly at TAF. This level is also used in the RPV CONTROL - NO ATWS procedure when all attempts to restore and maintain RPV water level above TAF have failed.
RPV water level instrumentation is referenced to the Top of Active Fuel. 0" TAF equates to water level at TAF. -30'TAF therefore means that RPV water level is 30" below TAF.
With regard to the various situations involving a loss of RPV water level indications, in general, the EOP action point "cannot be determined" relates to the condition where the operator has no idea where RPV water level is, or cannot determine by any means available that the RPV water level is above the point where adequate core cooling can be assured.
Potential Loss: 2. RPV level < 0" TAF OR CANNOT be determined Core submergence is the mechanism of core cooling whereby each fuel element is completely covered with water. Indicated RPV water level at or above the top of active fuel (0" TAF) provides direct confirmation that adequate core cooling exists. Assurance of continued adequate core cooling through core submergence is achieved when RPV water level can be maintained at or above TAF. If RPV water level cannot be restored and maintained above the top of active fuel, less desirable means of assuring adequate core cooling must be employed, posing a possible threat to fuel clad barrier integrity.
Page 50 of 122 Revision Oc
1*Inn I*T*I*o r nlv efr rrpaLV NivelAr QJafinIn Annav IVDInn NIuOllo*aý Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER Fuel Clad barrier potential loss FC.2 is the same as RCS barrier loss RCS.2. Thus, this threshold is both a loss of the RCS barrier and a potential loss of the Fuel Clad barrier, appropriately escalating the emergency class to a Site Area Emergency classification.
RPV water level instrumentation is referenced to the Top of Active Fuel. 0" TAF equates to water level at TAF.
Even if a procedure instructs deliberately lowering RPV water level below the top of active fuel, a classification is warranted due to the potential for fuel damage under such extreme conditions.
With regard to the various situations involving a loss of RPV water level indications, in general, the EOP action point "cannot be determined" relates to the condition where the operator has no idea where RPV water level is, or cannot determine by any means available that the RPV water level is above the point where adequate core cooling can be assured.
Differences from NUMARC/NESP-007:
- 1. Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level. RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
References:
- 1. 2000-GLN-3200.01, Plant Specific Technical Guideline, p. C1-2
- 2. 2000-BAS-3200.02, EOP Users Guide
- b. Drrwell Radiation Monitoring Loss: 1. ContainmentHi-Range RadiationMonitoringSystem (CHRRMS) > 440 R/hr7]
The CHRRMS reading indicates the release into the drywell of reactor coolant with elevated activity indicative of fuel damage. The reading assumes the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 pitCi/gm dose equivalent 1-131 into the drywell atmosphere.
Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage (approximately 2 - 5% clad failure depending on core inventory and RCS volume). This value is higher than that specified for RCS barrier Loss RCS.4; thus, this threshold indicates a loss of both Fuel Clad barrier and RCS barrier.
Page 51 of 122 Revision 0c
"17 v*h,*n *Tn,.l,* *-
Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER It is important to note that the specified value is only applicable under LOCA conditions.
Since the drywell CHRRMS may also be sensitive to shine from the RPV and piping, elevated readings are to be expected for conditions where fuel damage has occurred but there has been no release of coolant into the drywell atmosphere. It is important to recognize that, in the event the radiation monitor is sensitive to shine from the RPV or piping, spurious readings may be present and another indicator of fuel clad damage may be necessary.
I Potential Loss: None Not Applicable Differences from NUMARCINESP-007: None.
References:
- 1. Rad Engineering Calculation No. 2820-99-017
- c. Drmwell Pressure Not Applicable
- d. Breached / Bypassed (Primary Coolant Activity Level)
Loss: 1. Coolantactivity exceeds 300,uCi/gm (DEI) I This amount of coolant activity is well above that expected for iodine spikes and corresponds to about 2% to 5% fuel clad damage, indicating significant clad heating and thus the Fuel Clad Barrier is considered lost.
Page 52 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER Potential Loss: None.
Not Applicable Differences from NUMARC/NEI007: None.
References:
- 1. Rad Engineering Calculation No. 2820-99-012
- 2. Rad Engineering Calculation No. 2820-99-017
- 3. Rad Engineering Calculation No.96-004
- e. Containment Hydrogen Concentration Not Applicable
- f. Emergency Director Judgment Loss and Potential Loss: 1. Any condition in thejudgment of the Emergency Director that indicatesLoss or PotentialLoss of the Fuel Clad barrier This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost or potentially lost. The inability to monitor the barrier (e.g., RPV water level cannot be determined in the EOPs, etc.) should also be a factor in Emergency Director judgment that the barrier may be lost or potentially lost.
Differences from NUMARC/NEI007: None.
- 1. 2000-PLN-1300.01, OCNGS Emergency Plan, section 1.1.22 Page 53 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- 2. RCS BARRIER BASES
- a. RPV Water Level Loss: 1. RPV level < 0" TAF (not intentionally lowered by procedure)OR CANNOT be determined This threshold addresses the potential concern of adequate core cooling implicitly resulting from major failure of plant functions needed for the protection of the public. It is based on the EOP concern that the only mechanism remaining to assure adequate core cooling is steam cooling.
RCS barrier loss RCS.2 is the same as the Fuel Clad barrier potential loss FC.2. Thus, this threshold is both a loss of the RCS barrier and a potential loss of the Fuel Clad barrier, appropriately escalating the emergency class to a Site Area Emergency classification.
RPV water level instrumentation is referenced to the Top of Active Fuel. 0" TAF equates to water level at TAF.
With regard to the various situations involving a loss of RPV water level indications, in general, the EOP action point "cannot be determined" relates to the condition where the operator has no idea where RPV water level is, or cannot determine by any means available that the RPV water level is above the point where adequate core cooling can be assured If the EOPs instruct deliberately lowering RPV water level below the top of active fuel under ATWS conditions, the RCS is not assumed to be lost or challenged as a result.
Potential Loss: None.
Not Applicable Differences from NUMARC/NEI007:
- 1. Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine RPV water level. RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
References:
- 1. 2000-GLN-3200.01, Plant Specific Technical Guideline, p. Cl-2
- 2. 2000-BAS-3200.02, EOP Users Guide Page 54 of 122 Revision 0c
17v*lt*r. *T1, *.1,*,* *..
nj g Cra.lr NuupI~kl, d fk tnI A "una, -LVrlI IV U .Ina.l Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- b. Drywell Radiation Monitoring Loss: 1. ContainmentHi-Range RadiationMonitoringSystem (CHRRMS) > 45 R/hr The CHRRMS reading is a value that indicates the release of reactor coolant with coolant activity at the Technical Specification activity limit into the drywell. The reading corresponds to the Hi alarm set point on RE 790 and 791. This value also initiates closure of Torus/DW vent and purge isolation valves V-27-1, V-27-2, V-27-3, V-27-4, V-28-17, and V-28-18. This threshold is less than that specified for Fuel Clad barrier FC.3; thus, it is indicative of a RCS leak only. If the radiation monitor reading increases to the value specified by FC.3, fuel damage would also be indicated requiring declaration of a Site Area Emergency.
Potential Loss: None. 7 Not Applicable Differences from NUMARC/NEI007: None.
References:
- 1. 2000-RAP-3024.01 NSSS Alarm Response Procedures, 10-F-4-K
- 2. Rad Engineering Calculation No. 2820-99-017
- c. Drywell Pressure Loss: 1.Drywellpressure>3.0psigAND indicationof a RCS leak inside drywell Drywell pressure in excess of the drywell high pressure scram setpoint is designed to be indicative of a LOCA event. The phrase "and indication of a RCS leak inside drywell" has been added to exclude drywell pressurization events that are not caused by a loss of the RCS barrier (e.g., extended loss of drywell cooling). If this threshold is exceeded, there is a clear indication that a leak of sufficient magnitude exists that prevents drywell pressure stabilization.
Potential Loss: None.
Not Applicable Differences from NUMARC/NEI007:
Page 55 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- 1. The NUMARC EAL contains only the drywell pressure threshold. The qualifying condition "and indication of a RCS leak inside drywell" has been added as a human factors reminder to the Emergency Director that this EAL is for accident scenarios only. Thus, a drywell pressure increase due to the loss of drywell cooling will not require an emergency classification.
References:
- 1. 2000-BAS-3200.02, EOP User's Guide, p. 2-2
- d. Breached /Bypassed Loss: 1. UnisolableMain Steam Line breakoutside containment OR
- 2. UnisolableIsolation Condenser tube rupture Unisolable infers that the leak that cannot be isolated from the Control Room.
When evaluating this EAL for unisolable primary system leakage, it is appropriate to attempt isolation from the Control Room prior to classification.
Unisolable Main Steam Line break is not meant to cause a declaration based on leaks such as valve packing leaks where the consequences offsite would be negligible.
Unisolable Isolation Condenser tube rupture is meant to be an unisolable condenser tube rupture indicative of> 50 gpm primary system leakage.
Potential Loss: 3. RCS leakage >50 gpm OR
- 4. Unisolableprimary system leakage outside of drywell as indicated by exceeding EITHER of the following in one or more areas requiringa scram:
EMG-3200.11 Max Normal Temperature OR EMG-3200.11 Max NormalRadiationLevel Unisolable infers that the leak that cannot be isolated from the Control Room.
When evaluating this EAL for unisolable primary system leakage, it is appropriate to attempt isolation from the Control Room prior to classification.
The potential loss of RCS based on leakage is set at a level indicative of a small breach of the RCS but which is well within the makeup capability of normal and emergency high pressure systems. Core uncovery is not a significant concern for a 50 gpm leak; however, Page 56 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER a break propagation leading to a significantly larger loss of inventory is possible. RCS leakage is measured by the normal primary system leakage monitoring system and is leakage into the drywell. Under certain conditions, this system may be isolated due to increased drywell pressure caused by the leak. In that case, a "loss" of RCS will be indicated and this "potential loss" of RCS would not impact the classification.
Inventory loss events, such as a stuck open Electro-Mechanical Relief Valve (EMRV),
should not be considered when referring to "RCS leakage" because they are not indications of a break, which could propagate.
Potential loss of RCS based on primary system leakage outside the drywell is determined from secondary containment area temperatures or radiation levels. EOP guidance stipulates that when the secondary containment temperature or radiation maximum normal value has been exceeded for one area, all systems, except those required for EOP actions or fire suppression, be isolated. The reactor may be manually scrammed if the high temperature or radiation level continues to increase and is being caused by an unisolable primary system discharge into the reactor building. Therefore, it is appropriate to direct emergency classification based on elevated secondary containment temperature and radiation levels.
Secondary containment areas and maximum normal operating temperatures and radiation levels are given in EMG-3200.11.
Differences from NUMARC/NEI007: None.
References:
- 1. 2000-GLN-3200.01, Plant Specific Technical Guideline, p. SC-6
- 2. EMG-3200.1 1, Secondary Containment Control
- e. Containment Hydrogen Concentration Not Applicable
- f. Emeruency Director Jud2ment Loss and Potential Loss: 1. ANY condition in thejudgment ofthe Emergency Director that indicatesLoss or PotentialLoss of the RCS barrier This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost or potentially lost. The inability to monitor the barrier (e.g., RPV water level cannot be determined in the EOPs, etc.) should also be a factor in Emergency Director judgment that the barrier may be lost or potentially lost.
Differences from NUMARC/NEI007: None.
References:
- 1. 2000-PLN-1300.01, OCNGS Emergency Plan, section 1.1.22 Page 57 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- 3. PRIMARY CONTAINMENT BARRIER BASES
- a. RPV Water Level I Loss: None Not Applicable Potential Loss: 1. Entry into SAMGs as requiredby EOPs Entry to the Severe Accident Management Guidelines is prescribed by the EOPs as follows:
"* RPV water level cannot be restored and maintained above the Minimum Steam Cooling RPV Water Level (MSCRWL) (EMG-3200.01A RPV Control -No ATWS or EMG-3200.O1B RPV Control - With ATWS)
"* RPV water level cannot be determined and core damage is occurring (EMG 3200.08A RPV Flooding - No ATWS or EMG-3200.08B RPV Flooding - With ATWS)
"* Drywell or torus hydrogen concentration reaches 2.5%
These conditions represent imminent melt sequences which, if not corrected, could lead to vessel failure and increased potential for containment failure. In conjunction with the RPV water level Fuel Clad and RCS barrier thresholds, this Containment potential loss results in the declaration of a General Emergency (loss of two barriers and the potential loss of a third).
Differences from NUMARC/NEI007:
- 1. The Revision 2 NUMARC EAL prescribes an RPV water level in conjunction with the Maximum Core Uncovery Time Limit (MCUTL). This is a misapplication of the MCUTL, which was corrected in revision 3 of NUMARC/NESP-007. Primary Containment Flooding required (entry into SAMG) is now specified in the current NUMARC document.
References:
- 5. EMG-3200.02 - Primary Containment Control Page 58 of 122 Revision 0c
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- FISSION PRODUCT BARRIER
- b. Drywell Radiation Monitoring Loss: None.
Not Applicable Potential Loss: 1. ContainmentHi-Range RadiationMonitoringSystem (CHRRMS)
> 2.OE+4Rrhr The CHRRMS reading is a value that indicates significant fuel damage (> 20% clad failures) well in excess of that required for loss of RCS and Fuel Clad barriers. This value assumes 20% clad failures with the subsequent release of RCS volume into the containment. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant. Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, warranting declaration of a General Emergency. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%.
Differences from NUMARCiNEI007: None.
References:
- 1. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents
- 2. EMG-3200.02, Primary Containment Control
- 3. Rad Engineering Calculation No. 2820-99-017
- c. Drywell Pressure Loss: 1. Rapid,unexplaineddrop in drywell pressurefollowing an initialrise OR
- 2. Drywellpressure response not consistent with LOCA conditions indicatinga Containment breach Rapid unexplained loss of pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of containment integrity.
Drywell pressure should increase as a result of mass and energy release into containment from a LOCA. In a design-basis LOCA event, drywell pressure is expected to reach 38.1 psig. Thus, drywell pressure not increasing under these conditions indicates a loss of containment integrity.
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FISSION PRODUCT BARRIER Potential Loss: 3. Drywellpressure> 44psig I The threshold pressure is the FSAR drywell design pressure at 292F.
Differences from NUMARC/NEI007: None.
References:
- 1. OCNGS Technical Specifications section 5.2 Basis (38.1 psig)
- 3. EMG-3200.02, Primary Containment Control
- d. Breached / Bypassed Loss: 1. FailureofALL isolation valves in any one linepenetratingPrimary Containment to close resultingfrom an isolation actuationsignal when required AND Downstream pathway exists to environment This threshold addresses containment isolation failures allowing a direct flow path to the environment such as failure of both MSIVs to close following a Main Steam Line break or when an isolation is required with open valves downstream to the turbine or to the condenser.
Loss: 2. Intentionalventing per EMG-3200.02 is requiredwith drywellpressure
> 3.0 psig Intentional venting of primary containment per the EOPs to the secondary containment and/or the environment is considered a loss of containment. EMG-3200.02, Primary Containment Control, specifies primary containment venting in Step PC/P-3 (for Primary Containment Pressure Limit) and Step PC/G-2 (for detectable hydrogen). This EAL threshold does not apply to venting of Primary Containment as needed to maintain pressure below the high drywell pressure setpoint (3.0 psig).
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FISSION PRODUCT BARRIER Loss: 3. Unisolableprimarysystem leakage outside of drywell as indicatedby exceeding EITHER of thefollowing in one or more areas requiringa scram:
EMG-3200.11 Max Normal Temperature OR EMG-3200.11 Max Normal RadiationLevel Unisolable infers that the leak that cannot be isolated from the Control Room.
When evaluating this threshold for unisolable primary system leakage, it is appropriate to attempt isolation from the Control Room prior to classification.
Potential loss of RCS based on primary system leakage outside the drywell is determined from EOP area temperatures or radiation levels. EOP guidance stipulates that when the secondary containment temperature or radiation maximum normal value has been exceeded for one area, all systems, except those required for EOP actions or fire suppression, be isolated. The reactor may be manually scrammed if the high temperature or radiation level continues to increase and is being caused by an unisolable primary system discharge into the reactor building. Therefore, it is appropriate to direct emergency classification based on elevated secondary containment temperature and radiation levels.
Secondary containment areas and maximum normal operating temperatures and radiation levels are given in EMG-3200.1 1.
Potential Loss: None. ]
Not Applicable Differences from NUMARC/NEI007:
- 1. Expanded the isolation failure of primary containment isolation valves to include lines without automatic isolation by deleting "resulting from an isolation actuation signal." Failures such as feedwater line break outside primary containment with failure of the check valve to fully close deserve classification as primary containment losses.
References:
- 1. EMG-3200.02, Primary Containment Control
- 2. 2000-GLN-3200.01, Plant Specific Technical Guideline, p. SC-6
- 3. EMG-3200.1 1, Secondary Containment Control Page 61 of 122 Revision 0e
n li*ti t'rDD
, V u1i.-l,*aAi l A .It VIATII f-lf N.TlS1n0 ll .11 Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- e. Containment Hydrogen Concentration Loss: None Not Applicable Potential Loss: 1. ContainmentH2 concentration>_6%
AND Drywell or tonis 02 concentration>_5%
The specified value of 6% hydrogen concentration is the minimum that 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.
This threshold is intended to cover situations in which the hydrogen production is due to the zirconium-water reaction expected in fuel melt sequences. The oxygen component may be achieved through venting the containment or other means are possible. Since the fuel clad must be breached to sustain the a zirconium-water reaction and the RCS must be breached to accumulate high hydrogen concentrations in containment, the threshold is a loss of 2 out of 3 fission product barriers with a potential loss (or actual loss) of the third.
If drywell or torus hydrogen concentration reaches 2.5 %, primary containment flooding is required, directing entry to the SAMGs. The presence of hydrogen concentrations in the deflagration range (6%) is therefore indicative of a severe accident condition.
Differences from NUMARC/NEI007: None.
References:
- 1. EMG-3200.02, Primary Containment Control
- f. Emergency Director Judgment Loss and Potential Loss: 1. ANY condition in the judgment of the Emergency Director that indicatesLoss or PotentialLoss ofthe Primary Containmentbarrier This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Primary Containment barrier is lost or potentially lost. The inability to monitor the barrier (e.g., RPV water level cannot be determined in the EOPs, Page 62 of 122 Revision 0c
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FISSION PRODUCT BARRIER etc.) should also be a factor in Emergency Director judgment that the barrier may be lost or potentially lost.
Differences from NUMARC/NESP-007: None.
References:
- 1. 2000-PLN-1300.01, OCNGS Emergency Plan, section 1.1.22 Page 63 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS System Malfunctions Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They are based upon the potential to pose actual or potential threats to plant safety.
The events of this category have been grouped into the following subcategories:
Loss of AC Power Loss of vital plant AC electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems that may be necessary to ensure fission product barrier integrity. This category includes losses of onsite and/or offsite AC power sources including station blackout events.
Loss of DC Power Loss of vital plant DC electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems that may be necessary to ensure fission product barrier integrity. This category involves total losses of vital plant 125 vdc power sources.
Failure of Reactor Protection System The inability to control reactor power below certain levels can pose a direct threat to reactor fuel, RPV and primary containment integrity.
Decay Heat Removal This subcategory includes events that are indicative of losses of operability of safety systems such as Residual heat Removal, or cold and hot shutdown capabilities.
Loss of Annunciators Certain events that degrade plant operator ability to effectively assess plant conditions warrant emergency classification. Losses of annunciators are in this subcategory.
RCS Leakage/RPV Draindown This subcategory includes events that are indicative of RCS leakage in excess of levels that may be indicative potential RCS breach as well as abnormally low RPV water levels.
Loss of Communication Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
Losses of communication equipment are in this subcategory.
Technical Specifications Only one EAL falls into this subcategory. It is related to the failure of the plant to be brought to the required plant operating condition required by technical specifications.
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SYSTEM MALFUNCTIONS MG1 INITIATING CONDITION Prolonged Loss of ALL Offsite AC Power AND Prolonged Loss of ALL Onsite AC Power EAL THRESHOLD VALUE BOTH 4160V Busses 1C and 1D de-energized for > 15 min.
AND ANY of the following:
"* Restoration of at least one emergency bus within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is not likely
"* RPV level CANNOT be maintained > 0" TAF OR CANNOT be determined
"* Torus water temperature and RPV pressure exceeds the Heat Capacity Temperature Limit (Figure F, EMG-3200.02)
MODE APPLICABILITY, 1 and 2 BASIS' Loss of all AC power compromises all plant safety systems requiring electric power including ECCS, containment heat removal and the ultimate heat sink. Prolonged loss of all AC power will cause core uncovering and loss of containment integrity. The 1-hour interval to restore AC power is based on a station blackout coping analysis performed in conformance with 10 CFR 50.63 and Regulatory Guide 1.155, "Station Blackout."
Although this EAL may be viewed as redundant to the Fission Product Barrier Degradation EALs, its inclusion is necessary to better assure timely recognition and emergency response.
This EAL is specified to assure that in the unlikely event of a prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.
The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions.
In addition, under these conditions, fission product barrier monitoring capability may be degraded. Although it may be difficult to predict when power can be restored, it is necessary to give the Emergency Director a reasonable idea of how quickly he may need to declare a General Emergency based on two major considerations:
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SYSTEM MALFUNCTIONS MG1 - Cont'd BASIS - Cont'd
- 1. Are there any present indications that core cooling is already degraded to the point that loss or potential loss of fission product barriers is imminent?
- 2. If there are no present indications of such core cooling degradation, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of the third barrier can be prevented?
Thus, indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Director judgment as it relates to imminent loss or potential loss of fission product barriers and degraded ability to monitor fission product barriers.
RPV water level instrumentation is referenced to the Top of Active Fuel. 0" TAF equates to water level at TAF.
Core submergence is the mechanism of core cooling whereby each fuel element is completely covered with water. Indicated RPV water level at or above the top of active fuel (0" TAF) provides direct confirmation that adequate core cooling exists. Assurance of continued adequate core cooling through core submergence is achieved when RPV water level can be maintained at or above TAF. If RPV water level cannot be restored and maintained above the top of active fuel, less desirable means of assuring adequate core cooling must be employed, posing a possible threat to fuel clad barrier integrity.
Even if a procedure instructs deliberately lowering RPV water level below the top of active fuel, a classification is warranted due to the potential for fuel damage under such extreme conditions.
With regard to the various situations involving a loss of RPV water level indications, in general, the EOP action point "cannot be determined" relates to the condition where the operator has no idea where RPV water level is, or cannot determine by any means available that the RPV water level is above the point where adequate core cooling can be assured.
Torus water temperatures in excess of the Heat Capacity Temperature Limit (HCTL) is a distinct indication that heat removal from the primary containment is extremely challenged. The HCTL is a function of RPV pressure and torus water temperature. This limit defines the set of RPV pressure and torus water temperature combinations such that, should an emergency RPV depressurization be initiated from those conditions, the final torus water temperature will not result in pressurizing the primary containment above the Primary Containment Pressure Limit (PCPL).
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SYSTEM MALFUNCTIONS MG1 - Cont'd
ýBASIS - Cont'd Emergency Buses IC and ID can be powered from non-emergency Buses 1A and lB and emergency diesel generators EDG-1 and EDG-2.
Buses IA and 1B can be powered from the Auxiliary transformer, and Startup transformers SA and SB. Bus 1B can also be powered by the SBO transformer.
REhFERENCE(S)
- 1. EMG-3200.02, Primary Containment Control
- 2. 2000-BAS-3200.02, EOP User's Guide
- 3. 10CFR50.63
- 4. Regulatory Guide 1.155, Station Blackout
- 1. Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level. RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
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SYSTEM MALFUNCTIONS MS1 INITIATING CONDITION Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses EAL THRESHOLD VALUE BOTH 4160V Busses IC and 1D de-energized for > 15 min.
ýMODE APPLICABILITY 1 and 2 BASIS Loss of all AC power compromises all plant safety systems requiring electric power including ECCS, containment heat removal and the ultimate heat sink. Prolonged loss of all AC power will cause core uncovering and loss of containment integrity, thus this event can escalate to a General Emergency.
Emergency Buses IC and 1D can be powered from non-emergency Buses 1A and 1B and emergency diesel generators EDG-1 and EDG-2.
Buses IA and lB can be powered from the Auxiliary transformer, and Startup transformers SA and SB. Bus lB can also be powered by the SBO transformer.
REFERENCE(S) .
'DIFFERENCES None Page 68 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MAl
'INITIATING CONDITION AC power capability to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout
- EAL THRESHOLD VALUE Loss of offsite power to BOTH 4160V Busses 1C and ID for > 15 min.
AND EITHER of the 4160V busses 1C or 1D de-energized for > 15 min.
MODE APiPLICABILITY 1 and 2
'BASIS This EAL is intended to provide escalation from the loss of AC power Unusual Event EAL. The condition of this EAL could occur due to:
" Loss of all offsite power concurrent with a failure of one emergency diesel generator to supply its emergency bus.
" Loss of both emergency diesel generators with only one emergency bus being powered from a single available source:
- Bus IC from Bus 1A fed from either Startup transformer SA OR Auxiliary transformer OR
- Bus 1D from Bus 1B fed from either Startup transformer SB OR Auxiliary transformer OR the SBO Transformer.
Any additional single failure would result in a station blackout.
Emergency Buses 1C and ID can be powered from non-emergency Buses 1A and lB and emergency diesel generators EDG-1 and EDG-2.
Buses 1A and 1B can be powered from the Auxiliary transformer and Startup transformers SA and SB. Bus 1B can also be powered by the SBO transformer.
REFERENCE(S)
- 1. OCNGS Drawing BR 3000
- 2. 2000-ABN-3200.37 Station Blackout section 1.0 Page 69 of 122 Revision 0c
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SYSTEM MALFUNCTIONS MAI - Cont'd "NUMARCIC SA5 DIFFERENCES None Page 70 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MA2
- INITIATING CONDITION Loss of All Offsite Power AND Loss of All Onsite AC Power to Essential Busses During Cold Shutdown Or Refueling Mode EAL THRESHOLD VALUE BOTH 4160V Busses 1C and ID de-energized for > 15 min.
MODE APPLiCABILITY 3,4 andD BASIS Loss of all AC power compromises all plant safety systems requiring electric power including ECCS, containment heat removal, spent fuel heat removal and the ultimate heat sink. When in cold shutdown, refueling, or defueled mode, the event can be classified as an Alert because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL. Escalating to Site Area Emergency, if appropriate, is by Abnormal Rad Levels/Radiological Effluent, or Emergency Director Judgment. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Emergency Buses IC and 1D can be powered from non-emergency Buses 1A and 1B and emergency diesel generators EDG-l and EDG-2.
Buses 1A and lB can be powered from the Auxiliary transformer and Startup transformers SA and SB. Bus 1B can also be powered by the SBO transformer.
(S)
REFER&ENcE
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MUl INITIATING CONDITION Loss of All Offsite Power to Essential Busses for Greater Than 15 Min.
EAL THRESHOLD VALUE Loss of offsite power to BOTH 4160V Busses 1C and 1D for > 15 min.
MODE A&PPLICABiLITY All BASIS Prolonged loss of offsite AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete Loss of AC Power (Station Blackout). The fifteen-minute interval excludes emergency declaration due to transient or momentary power losses.
Emergency Buses IC and 1D can be powered from non-emergency Buses 1A and 1B and emergency diesel generators EDG-1 and EDG-2.
Buses 1A and lB can be powered from the Auxiliary transformer and Startup transformers SA and SB. Bus lB can also be powered by the SBO transformer.
'REFERENCE(S)
- Y*lnn N.eloar Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MS3 INITIATING CONDITION Loss of All Vital DC Power EAL THRESHOLD VALUE Loss of ALL vital DC power indicated by < 115 VDC indication on 125 VDC Busses B and C for> 15 min.
,-MODE APPLICABILITY 1 and 2 BASIS Loss of all DC power compromises ability to monitor and control plant safety functions.
Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is significant decay heat and sensible heat in the RPV. Escalation to a General Emergency would occur by Abnormal Rad Levels/Radiological Effluent, Fission Product Barrier Degradation, or Emergency Director judgment. The fifteen minute interval was selected as a threshold to exclude transient or momentary power losses.
OCNS has three 125 VDC electrical busses. Bus 'A' only powers non-vital equipment and therefore is not considered in the loss for this threshold. The specified bus voltage indication of 115 VDC is based on the minimum bus voltage necessary for the operation of Core Spray Pumps and incorporates a margin of at least 15 minutes of operation before the onset of inability to operate motor loads.
REFERENCE(S),
- 1. 2000-ABN-3200.13A/B, Loss of DC Distribution Center A and/or B
- 2. 2000-ABN-3200.13C, Loss of DC Distribution Center
- 3. ARP 2000-RAP-3024.02
- NUMARC IC SS3
'DIFFERENCES None Page 73 of 122 Revision 0c
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SYSTEM MALFUNCTIONS MU3
'INITIATING CONDITION Unplanned Loss of Required DC Power During Cold Shutdown or Refueling Mode > 15 Min.
EAL THRESHOLD VALUE Loss of ALL vital DC power indicated by < 115 VDC indication on 125 VDC Busses B and C for > 15 min.
MODE APPLICABILITY 3 and 4 JBASIS The purpose of this EAL is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss.
Unplanned is included in this IC and EAL to preclude the declaration of an emergency as a result of planned maintenance activities. Routinely, maintenance on a DC distribution center is performed during shutdown periods. It is intended that the loss of the operating (operable) distribution centers are to be considered. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert is required by another EAL.
OCNS has three 125 VDC electrical busses. Bus 'A' only powers non-vital equipment and therefore is not considered in the loss for this threshold. The specified bus voltage indication of 115 VDC is based on the minimum bus voltage necessary for the operation of Core Spray Pumps and incorporates a margin of at least 15 minutes of operation before the onset of inability to operate motor loads.
REFERENCE(S)
- 1. 2000-ABN-3200.13A/B, Loss of DC Distribution Center A and/or B
- 2. 2000-ABN-3200.13C, Loss of DC Distribution Center
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MG4 "INITIATINGCONDITION Auto and manual SCRAM NOT successful AND loss of core cooling or heat sink
,EAL THRESHOLD VALUE RPS setpoint for an automatic SCRAM exceeded AND Failure of automatic RPS, ARI and manual SCRAM to reduce reactor power < 2%
AND EITHER:
RPV level CANNOT be restored and maintained > -30" TAF OR CANNOT be determined OR Torus water temperature and RPV pressure exceeds the Heat Capacity Temperature Limit (Figure F, EMG-3200.02)
MODE APPLICABILITY 1
BASIS A valid automatic and/or manual scram signal is present as indicted by control room indications and/or alarms and APRM indication at or above 2% power. The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automatically). The system is "fail safe" meaning it de-energizes to function.
An Anticipated Transient Without Scram (ATWS) event can be caused either by a failure of RPS (electrical failure) or a failure of the Control Rod Drive system to permit the control rods to insert (hydraulic failure).
A failure of the Reactor Protection System to sufficiently shut down the reactor (as indicated by reactor power remaining at or above 2%) is a degraded plant condition that together with suppression pool temperature approaching the Boron Injection Initiation Temperature requires the injection of boron to shut down the reactor. With Reactor Power less than 2% the heat being generated in the core can be removed from the RPV and containment while actions are taken to bring the reactor subcritical.
A manual scram is defined as any set of actions by the reactor operator(s) at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical. Taking the mode switch to shutdown as part of the actions required by the EOPs is considered a manual scram action.
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1vnInii NuuplDol.
fl .L efor rrccL- N"Anav Qfaf;riin Ainnow Vvalnn TT"olý Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MG4 - Cont'd
,BASIS - Cont'd This EAL is not applicable if a manual scram is initiated and no RPS setpoints are exceeded. Taking the mode switch to shutdown is considered a manual scram action.
RPV water level instrumentation is referenced to the Top of Active Fuel. 0" TAF equates to water level at TAF.
The RPV water level is the Minimum Steam Cooling RPV Water Level (MSCRWL) and is used in EOPs to indicate challenge to core cooling. The MSCRWL is the lowest RPV water level at which the submerged portion of the reactor core will generate sufficient steam to prevent any clad in the uncovered portion of the core from heating to 1500'F; the threshold temperature of fuel clad perforation. This water level is utilized to preclude fuel damage when RPV water level is below the top of active fuel (TMF). The MSCRWL appears in the RPV CONTROL - WITH ATWS EOP when RPV water level is intentionally lowered to reduce reactor power. When RPV water level is deliberately lowered, power instabilities may produce noticeable oscillations in RPV water level and make it difficult to maintain water level exactly at TAF.
With regard to the various situations involving a loss of RPV water level indications, in general, the EOP action point "cannot be determined" relates to the condition where the operator has no idea where RPV water level is, or cannot determine by any means available that the RPV water level is above the point where adequate core cooling can be assured.
Torus water temperatures in excess of the Heat Capacity Temperature Limit (HCTL) is a distinct indication that heat removal from the primary containment is extremely challenged. The HCTL is a function of RPV pressure and torus water temperature. This limit defines the set of RPV pressure and torus water temperature combinations such that, should an emergency RPV depressurization be initiated from those conditions, the final torus water temperature will not result in pressurizing the primary containment above the Primary Containment Pressure Limit (PCPL).
REFERENCE(S)
- 1. 2000-BAS-3200.02, EOP User's Guide
- 2. EMG-3200.02, Primary Containment Control NUMARC IC SG2 Page 76 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MG4 - Cont'd JDIFFERENCES Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level. RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
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Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MS4 INITIATING CONDITION Auto and manual SCRAM NOT successful EAL THRESHOLD VALUE RPS setpoint for an automatic SCRAM exceeded AND Failure of automatic RPS, ARI and manual SCRAM to reduce reactor power < 2%
MODE APPLICABILITY BASiS A valid automatic and/or manual scram signal is present as indicted by control room indications and/or alarms and APRM indication at or above 2% power. The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automatically). The system is "fail safe," that is, it de-energizes to function.
An Anticipated Transient Without Scram (ATWS) event can be caused either by a failure of RPS (electrical failure) or a failure of the Control Rod Drive system to permit the control rods to insert (hydraulic failure).
A failure of the Reactor Protection System to sufficiently shut down the reactor (as indicated by reactor power remaining at or above 2%) is a degraded plant condition that together with torus water temperature approaching the Boron Injection Initiation Temperature requires the injection of boron to shut down the reactor. With Reactor Power less than 2% the heat being generated in the core can be removed from the RPV and containment while actions are taken to bring the reactor subcritical.
A manual scram is defined as any set of actions by the reactor operator(s) at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical. Taking the mode switch to shutdown as part of the actions required by the EOPs is considered a manual scram action.
This EAL is not applicable if a manual scram is initiated and no RPS setpoints are exceeded. Taking the mode switch to shutdown is considered a manual scram action.
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Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MS4 - Cont'd REFERENCE(S)
- 1. 2000-BAS-3200.02, EOP User's Guide
,NUMARC IC SS2 DIFFERENCES None Page 79 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MA4 INITIATING CONDITION Auto SCRAM NOT successful OR Loss of Manual SCRAM Capability EAL THRESHOLD VALUE' EITHER:
- 1. RPS setpoint for an automatic SCRAM exceeded AND Failure of automatic SCRAM to achieve reactor shutdown OR
- 2. Loss of manual SCRAM capability indicated by failure of ALL manual SCRAM attempts to achieve reactor shutdown MODE APPLICABILITY BASIS Condition (1) indicates failure of the automatic protection system to scram the reactor.
This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient and thus the plant safety has been compromised, and design limits of the fuel may have been exceeded. An Alert is indicated because conditions exist that lead to potential loss of fuel clad or RCS. Reactor protection system setpoint being exceeded (rather than limiting safety system setpoint being exceeded) is specified here because failure of the automatic protection system is the issue. Failure of manual scram would escalate the event to a Site Area Emergency.
'Reactor Shutdown' is defined to mean the reactor is sub-critical with reactor power below the heating range.
Condition (2) indicates failure of all manual SCRAM capability. While failure of all manual SCRAM capability does not challenge fuel design limits, it is indicative of a condition in which rapid reactor shutdown cannot be established prior to the fuel being challenged should an RPS setpoint subsequently be exceeded.
A manual scram is any set of actions by the reactor operator(s) at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical, including manual scram buttons, Mode Switch and actuation of ARI.
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n + w~fa CriaIr NWuu~1ar t~~~~~Fwii 4Z~afmn"E ~tl A nnow fltpf .lotaCA vIrNurir T Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MA4 con't
'REFERENCE(S)
- 1. 2000-BAS-3200.02, EOP User's Guide
- 1. Failure of manual SCRAM capability has been added to address conditions in which no RPS setpoint has been exceeded but all means of manual SCRAM have failed.
This additional threshold, not specified by NUMARC, was deemed appropriate to be anticipatory to failure of automatic scram signals as a result of exceeding an RPS setpoint.
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Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MS5 INITIATING CONDITION Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown
,EAL THRESHOLD VALUE Torus water temperature and RPV pressure CANNOT be maintained below the Heat Capacity Temperature Limit (Figure F, EMG-3200.02)
'MODE APPLICABILITY 1 and 2 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 addressed by EMG 3200.02 torus water temperature leg requiring an Emergency RPV Depressurization when parameters cannot be maintained below the Heat Capacity Temperature Limit (HCTL).
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 judgment, or fission product barrier degradation.
REFERENCE(S)
- 1. 2000-BAS-3200.02, EOP User's Guide
- 2. EMG-3200.02, Primary Containment Control NUMARC IC SS4 DIFFERENCES
- 1. Implements NUMARC/NESP-007 Rev. 3 BWR specific criteria. Revision 2 of NUMARC/NESP-007 simply specified loss of [site-specific function] necessary to maintain Hot Shutdown. Revision 3 of NUMARC is specific in defining this condition for BWRs as inability to maintain parameters below Heat Capacity Temperature Limit.
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Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MA5 INITIATING CONDITION Inability to Maintain Plant in Cold Shutdown EAL THRESHOLD VALUE Unplanned loss of all Technical Specification required systems available to provide decay heat removal functions AND Uncontrolled temperature rise that approaches or exceeds 212OF MODE APPLICABILITY 3 and 4 BASIS This EAL addresses complete loss of functions required for core cooling during refueling and cold shutdown modes. "Uncontrolled" means that system temperature increase is not the result of planned actions by the plant staff.
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.
Technical Specification required systems include: RHR, IC, CRD, Service Water, RBCCW and Core Spray Systems as well as offsite electrical power transformers and Emergency and SBO Diesel Generators.
1REFERENCE(S)
- 1. OCNGS Technical Specifications definitions section 1.7 NUMARC IC SA3 DIFFERENCES None Page 83 of 122 Revision Oc
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MS6 INITIATING CONDITION Inability to Monitor a Significant Transient in Progress EAL THRESHOLD VALUE A significant plant transient is in progress (Table M-1)
AND All of the following are lost:
- Safety system annunciators (Table M-2)
- Safety function indicators (Table M-3)
- Plant Process Computer Table M Significant Plant Transients
- > 25% thermal power change 0 Sustained power oscillations (30 watts/cm 2 LPRM peak to peak) 0 Stuck open EMRVs 0 ECCS Injections Table M Safety System Annunciators 0 ECCS(B,C)
- Containment Isolation (G, H, J) 0 Reactor Scram (G)
- Process Radiation Monitoring (1 OF)
Table M Safety Function Indicators
- Reactor Level, Pressure and Power (Panel 4F, 5F, 6F)
- Decay Heat Removal (Panel 1F/2F)
- Containment Safety Functions (Panel 1 IF, 12XR, 16R)
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Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MS6 - Cont'd
- MODE APPLICABILITY 1 and 2
ýBASIS This EAL is intended to recognize the inability of the control room staff to monitor the plant response to a transient. A Site Area Emergency is considered to exist if the control room staff cannot monitor safety functions needed for protection of the public.
Annunciators for this EAL should be limited to include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e.g.,
rad monitors, etc.)
The Plant Process Computer is not available to provide compensatory indication.
"Planned" actions are excluded from this EAL since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not an ameliorating factor.
"Significant transient" includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, level/pressure transients such as emergency RPV depressurization or ECCS injection, or reactor power oscillations of 10% or greater (> 30 watts/cm2 peak-to-peak).
Table M-2 lists those system annunciator panels considered to be safety related. Table M 3 lists those panel indications important for monitoring.
JREFERENCE(S)
- 1. None
,NUMARC IC SS6 DIFFERENCES None Page 85 of 122 Revision 0e
l*.Yalan Nnrloar Ovster Creek Nuclear Staltion Annex E'rplon NJuip-nr Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MA6 INITIATING CONDITION Unplanned Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a Significant Transient in Progress, or (2) Compensatory Non Alarming Indicators are Unavailable EAL THRESHOLD VALUE Unplanned loss, for > 15 min., of MOST (Note 1) or ALL of EITHER:
"* Safety system annunciators (Table M-2)
"* Safety function indicators (Table M-3)
AND EITHER:
"* A significant plant transient is in progress (Table M-1)
"* Plant Process Computer is unavailable Table M Significant Plant Transients
"* Scram
"* > 25% thermal power change
"* Sustained power oscillations (30 watts/cm 2 LPRM peak to peak)
"* Stuck open EMRVs
"* ECCS Injections Table M Safety System Annunciators
"* ECCS (B, C)
"* Containment Isolation (G, H, J)
"* Reactor Scram (G)
"* Process Radiation Monitoring (IOF)
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1 n,* vt un",.,n An wr . nLflf WIIný,t .,
Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MA6 - Cont'd EAL THRESHOLD VALUE - Cont'd Table M Safety Function Indicators
"* Reactor Level, Pressure and Power (Panel 4F, 5F, 6F)
"* Decay Heat Removal (Panel 1F/2F) 0 Containment Safety Functions (Panel 11F, 12XR, 16R)
"MODE APPLICABILITY 1 and 2 BASIS This EAL recognizes the difficulty associated in monitoring conditions without normal annunciators. In the opinion of the Shift Supervisor this loss of annunciators requires increased surveillance to safely operate the plant. "Most" refers to a loss of -75% or a significant risk that a degraded plant condition could go undetected. It is not intended that a detailed count of instrumentation be performed, but that only a rough approximation be used to determine the severity of the loss. The Plant Process Computer System is available to provide compensatory indication. Fifteen minutes is used as a threshold to exclude transient or momentary power losses. Unplanned loss of annunciators excludes scheduled maintenance and testing activities.
Table M-2 lists those system annunciator panels considered to be safety related. Table M 3 lists those indications important for monitoring.
It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50.72.
If the shutdown is not in compliance with the Technical Specification action, the Unusual Event is based on EAL 4.2.1 "Inability to Reach Required Shutdown Within Technical Specification Limits."
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
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Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MA6 - Cont'd
ýBASIS - Cont'd "Significant transient" includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, level/pressure transients such as emergency RPV depressurization or ECCS injection, or reactor power oscillations of 10% or greater (> 30 watts/cm 2 peak-to-peak).
Due to the limited number of safety systems in operation during cold shutdown, refueling, and defueled modes, no EAL is indicated during these modes of operation.
REFERENCE(S)
- 1. The condition "In the opinion of the Shift Supervisor, the loss of the annunciators or indicators requires increased surveillance to safely operate the unit(s)" has not been included in the condition consistent with changes to IC SA4 in NUMARC/NESP-007 Rev. 3. This statement does not provide useful assessment criteria to the EAL threshold.
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Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MU6 INITIATING CONDITION Unplanned Loss of Most or All Safety System Annunciation or Indication in the Control Room > 15 Min.
EAL THRESHOLD VALUE Unplanned loss, for > 15 min., of MOST (Note 1)or all of EITHER:
Safety system annunciators (Table M-2)
- Safety function indicators (Table M-3)
Table M Safety System Annunciators
"* ECCS(B,C)
"* Containment Isolation (G, H, J)
"* Reactor Scram (G)
"* Process Radiation Monitoring (1OF)
Table M Safety Function Indicators
"* Reactor Level, Pressure and Power (Panel 4F, 5F, 6F)
"* Decay Heat Removal (Panel 1F/2F)
"* Containment Safety Functions (Panel 11 F, 12XR, 16R)
MODE APPLICABILITY I and 2
- BASIS This EAL recognizes the difficulty associated in monitoring conditions without normal annunciators. In the opinion of the Shift Supervisor this loss of annunciators requires increased surveillance to safely operate the plant. "Most" refers to a loss of -75% or a significant risk that a degraded plant condition could go undetected. It is not intended that a detailed count of instrumentation be performed, but that only a rough approximation be used to determine the severity of the loss. The Plant Process Computer System is available to provide compensatory indication. Fifteen minutes is used as a threshold to exclude transient or momentary power losses. Unplanned loss of annunciators excludes scheduled maintenance and testing activities.
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Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MU6 - Cont'd BASIS - Cont'd Table M-2 lists those system annunciator panels considered to be safety related. Table M 3 lists those indications important for monitoring.
It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50.72.
The fifteen-minute interval was selected as a threshold to exclude transient or momentary power losses.
Due to the limited number of safety systems in operation during cold shutdown, refueling, and defueled modes, no EAL is indicated during these modes of operation.
REFERENCE(S)
- 1. 10 CFR 50.72
- 2. OCNS simulator walkdown VNUMARC IC SU3 DIFFERENCES
- 1. The condition "In the opinion of the Shift Supervisor, the loss of the annunciators or indicators requires increased surveillance to safely operate the unit(s)" has not been included in the condition consistent with changes to IC SU3 in NUMARC/NESP-007 Rev. 3. This statement does not provide useful assessment criteria to the EAL threshold.
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Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MS7 INITIATING CONDITION Loss of Water Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reactor Vessel EAL THRESHOLD VALUE RPV level < 0 " TAF OR CANNOT be determined MODE APPLICABILITY 3 and 4
'BASIS Under the condition specified by this EAL, severe core damage can occur and reactor coolant system pressure boundary integrity may not be assured. It is intended to address concerns raised by NRC Office for Analysis and Evaluation of Operational Data (AEOD)
Report AEODIEGO9, "BWR Operating Experience Involving Inadvertent Draining of the Reactor Vessel," dated August 8, 1986. This report states:
"In broadest terms, the dominant causes of inadvertent reactor vessel draining are related to the operational and design problems associated with the residual heat removal system when it is entering into or exiting from the shutdown cooling mode.
During this transitional period water is drawn from the reactor vessel, cooled by the residual heat removal system heat exchangers (from the cooling provided by the service water system), and returned to the reactor vessel. First, there are piping and valves in the residual heat removal system which are common to both the shutdown cooling mode and other modes of operation such as low pressure coolant injection and suppression pool cooling. These valves, when improperly positioned, provide a drain path for reactor coolant to flow from the reactor vessel to the suppression pool or the radwaste system. Second, establishing or exiting the shutdown cooling mode of operation is entirely manual, making such evolutions vulnerable to personnel and procedural errors. Third, there is no comprehensive valve interlock arrangement for all the residual heat removal system valves that could be activated during shutdown cooling. Collectively, these factors have contributed to the repetitive occurrences of the operational events involving the inadvertent draining of the reactor vessel."
Thus, declaration of a Site Area Emergency is warranted under the conditions specified by the EAL. Escalation to a general emergency is via radiological effluence IC RG1.
RPV water level instrumentation is referenced to the Top of Active Fuel. 0" TAF equates to water level at TAF.
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Oyster Creek Nuclear Station Annex Exelon Nuclear Nuclear yter C.... ... N cl.. ...
r ...tat......A.Exelon Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MS7 - Cont'd BASIS With regard to the various situations involving a loss of RPV water level indications, in general, the EOP action point "cannot be determined" relates to the condition where the operator has no idea where RPV water level is, or cannot determine by any means available that the RPV water level is above the point where adequate core cooling can be assured.
REFERENCE(S)..
- 1. 2000-GLN-3200.01, Plant Specific Technical Guideline, p. Cl-2
- 2. AEOD/EGO9, BWR Operating Experience Involving Inadvertent Draining of the Reactor Vessel
- 1. The condition stated in NUMARC NESP-007, SS5, L.a "Loss of all decay heat removal cooling as determined by (site-specific) procedure" is not necessary to conclude that the plant condition warrants a Site Area Emergency due to core uncovery; therefore, the example EAL was not included in this EAL.
- 2. Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level. RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
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Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MU7 INITIATING CONDITION RCS Leakage
ýEAL THRESHOLD VALUE Unidentified leakage > 10 gpm OR Identified leakage > 25 gpm MODE APPLICABILITY 1 and 2
`BASIS This EAL may be a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified leakage was selected as it is observable with normal control room indications. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances). The EAL for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.
Only operating modes in which there is fuel in the RPV and the RPV is pressurized are specified.
REFERENCE(S)
- 1. The value specified by NUMARC/NESP-007 Rev. 2 for unidentified leakage (10 gpm) is greater than that specified in the OCNS Technical Specifications. The greater value is utilized consistent with the NUMARC basis.
- 2. No reference is made to "pressure boundary leakage" since this term is not defined for OCNS and no distinction is made between "pressure boundary leakage" and "unidentified leakage" at OCNS.
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rvnlnn NTwur1in'ir n .Lcfar ranu Ininav Vvedn" WIVMAýý Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MU8 INITIATING CONDITION Unplanned Loss of All Onsite OR Offsite Communications Capabilities EAL THRESHOLD VALUE Loss of all onsite communications (Table M4) affecting the ability to perform routine operations OR Loss of all offsite communications (Table M-4)
Table M4 - Communications Onsite Offsite Plant Paging System X Conventional telephone lines X X Cell Phones X X Radio X X ERF X NRC Emergency Notification System (ENS) X Health Physics Network (HPN) X Bureau of Nuclear Engineering Information Line X ED Hotline X New Jersey State Police (NJSP Notification Line) X Ocean County Notification Line X NJ State ED Hotline X Environmental Assessment Direct Line X MODE APPLICABILITY All Page 94 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MU8 - Cont'd BASIS The purpose of this EAL is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72.
Onsite communications loss must encompass the loss of all means of routine communications.
Offsite communications loss must encompass the loss of all means of communications with offsite authorities. This EAL is intended to be used only when extraordinary means are being utilized to make communications possible (relaying of information from radio transmissions, individuals being sent to offsite locations, etc.). This loss is meant to include loss of the Meridian phone system, the Dedicated Telephone lines, the direct NJ Bell lines which are in the TSC, CR and OSC, the microwave lines and the radio channels between the site and the outside world. If notification can be accomplished via any of the above systems then conditions of the EAL are not met. On the other hand if conditions are met it will not be possible to make this notification from the site. It would be prudent to send a driver to a offsite location to attempt to complete the notification.
iREFERENcE(s). ..
- 1. 2000-PLN-1300.01 OCNGS Emergency Plan, section 7.4.1 and Table 12,
- p. E12-1
- 2. 10 CFR50.72 NUMARC IC SU6 DIFFERENCES None Page 95 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MU9 INITIATING CONDITION Plant is not brought to required operating mode within Technical Specifications LCO Action Statement Time EAL THRESHOLD VALUE Required operating mode is NOT reached within Tech. Spec. LCO action completion time "MODEAPPLICABILITY 1 and 2
ýBASIS Limiting Conditions of Operation (LCOs) require the plant to be brought to a required shutdown mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specifications requires NRC reporting under 10 CFR 50.72 (b) Non-emergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate notification of an Unusual Event is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an Unusual Event is based on the time at which the LCO specified action statement time period elapses under Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other System Malfunctions, Hazards, or Fission Product Barrier Degradation EALs.
REFERENCE(S)
- 1. OCNGS Technical Specifications
NUMARC IC SU2 DIFFERENCES None Page 96 of 122 Revision Oc
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS Hazards and Other Conditions Hazards are non-plant, system-related events that can directly or indirectly impact plant operation, reactor plant safety or personnel safety.
The events of this category have been grouped into the following subcategories:
Security Events This category includes unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.
Control Room Evacuation Conditions requiring evacuation of the Control Room due to fire, toxic gases or radiological conditions may affect plant operators ability to operate vital equipment.
Natural or Man-made Events Natural events include hurricanes, earthquakes or tornados that have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.
Man-made events are non-naturally occurring events that can cause damage to plant facilities and include turbine failures, aircraft/vehicle crashes or missile impacts.
Fire or Explosion Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the site Protected Area or which may affect operability of vital equipment.
Toxic or Flammable Gas Release Toxic or flammable gas releases can pose significant hazards to personnel and reactor safety particularly those which may restrict access to vital equipment.
Discretionary This category provides the Shift Manager or Emergency Director the latitude to use discretion in the declaration of emergencies based upon his/her judgment.
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Ovster Creek Nuclear Station Annex Exelnn Nncienr Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HG1 INITIATING CONDITION Security Event Resulting in Loss Of Ability to Reach AND Maintain Cold Shutdown EAL THRESHOLD VALUES
- 1. Loss of physical control of the Control Room due to a security event.
- 2. Loss of physical control of the remote shutdown capability due to a security event.
MODE APPLICABILITY ALL BASIS This class of security event represents conditions under which a hostile force has taken physical control of areas required to reach and maintain cold shutdown. Loss of Remote Shutdown Capability would occur if the control function of the Remote Shutdown Panels were lost.
Security events, which meet the threshold for declaration of a General Emergency, are physical loss of the Control Room or the Remote and Alternate Shutdown Panels.
This situation leaves the plant in a very unstable condition with a high potential of multiple barrier failures.
REFERENCE(S)
None NUMARC IC HG1 DIFFERENCES None Page 98 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HS1 INITIATING CONDITION Confirmed Security Event in a Vital Area
ýEAL THRESHOLD VALUES
- 1. Intrusion into plant Vital Area by a hostile force.
- 2. Confirmed bomb, sabotage or sabotage device discovered in a Vital Area.
MODE APPLICABILITY ALL BASIS This class of security event represents an escalated threat to plant safety above that contained in an Alert in that a hostile intrusion or attack has progressed from the Protected Area to a Vital Area. The Vital Areas are within the Protected Area and are generally controlled by key card readers. These areas contain vital equipment, which includes any equipment, system, device or material, the failure, destruction or release of could directly or indirectly endanger the public health and safety by exposure to radiation. Equipment or systems, which would be required to function to protect health and safety following such failure, destruction or release, are also considered vital.
This event will be escalated to a General Emergency based upon the loss of physical control of the Control Room or Remote Shutdown Capabilities.
REFERENCE(S)
None NUMARC IC HS1 DIFFERENCES None Page 99 of 122 Revision Oc
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HAI INITIATING CONDITION Confirmed Security Event in a Plant Protected Area EAL THRESHOLD VALUES
- 1. Intrusion into the Protected Area(s) by a hostile force.
- 2. Confirmed bomb, sabotage or sabotage device discovered in the Protected Area(s)
ýMODE APPLICABILITY ALL BASIS This class of security event represents an escalated threat to the level of safety of the plant.
This event is satisfied if physical evidence supporting the hostile intrusion or attack exists.
The Shift Management will declare an Alert subsequent after consulting with the on-shift Security representative to determine the validity of the entry conditions.
This event will be escalated to a Site Area Emergency based upon a hostile intrusion or act in-plant Vital Areas.
The Protected Areas for OCNS include both the Protected Area Boundary surrounding the plant process buildings and the Protected Area Boundary surrounding the Independent Spent Fuel Storage Installation.
REFERENCE(S)
None NUMARC IC HA4 DIFFERENCES None Page 100 of 122 Revision Oc
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HUI INITIATING CONDITION Confirmed Security Event That Indicates a Potential Degradation in the Level of Plant Safety EAL THRESHOLD VALUES
- 1. A credible threat to the station reported by the NRC.
- 2. BOTH of the following criteria are met for a credible threat reported by any other outside agency or determined per the Safeguards Contingency Plan:
"* Is specifically directed towards the station.
"* Is imminent (< 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />).
- 3. Attempted intrusion and attack of the Protected Area(s)
- 4. Attempted sabotage discovered within the Protected Area(s)
- 5. Hostage/Extortion situation that threatens normal plant operations MODE APPLICABILITY ALL BASIS A security threat that is identified as being directed towards the station and represents a potential degradation in the level of safety of the plant. A security threat is satisfied if physical evidence supporting the threat exists, if information independent from the actual threat exists, or if a specific group claims responsibility for the threat. The Shift Management will declare an Unusual Event subsequent to consulting with the on shift Security representative to determine the credibility of the security event per the Safeguards Contingency Plan.
The Protected Areas for OCNS include both the Protected Area Boundary surrounding the plant process buildings and the Protected Area Boundary surrounding the Independent Spent Fuel Storage Installation.
Security events which do not represent a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or 10 CFR 50.72 and will not cause an Unusual Event to be declared.
This event will be escalated to an Alert based upon a hostile intrusion or act within the Protected Area.
Page 101 of 122 Revision 0e
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HUI Cont'd REFERENCE(S)
'DIFFERENCES
- 1. This EAL threshold has been written to conform with IC HU4 regarding devices as amended and endorsed by the NRC in a letter from Mr. B. A. Boger to Ms. Lynette Hendricks (NEI) dated 2/4/02 Page 102 of 122 Revision 0c
Ovqte~r Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HS2 INITIATING CONDITION Control Room Evacuation Initiated AND Plant Control CANNOT be Established in < 15 min.
EAL THRESHOLD VALUES Control Room evacuation initiated AND Control of the plant CANNOT be established in < 15 min. per 2000-ABN-3200.30 "Control Room Evacuation" MODE APPLICABILITY:
All BASIS:
Control - Placing all local control switches in local control necessary for operation from remote panels and the Shift Manager has determined that the systems for controlling reactivity, core cooling and heat sink functions are established.
Transfer of safety system control has not been performed in an expeditious manner but it is unknown if any damage has occurred to the fission product barriers. The 15 minute time limit for transfer of control is based on a reasonable time period for personnel to leave the control room, arrive at the remote shutdown area, and reestablish plant control to preclude core uncovery and/or core damage. During this transitional period the function of monitoring and/or controlling parameters necessary for plant safety may not be occurring and as a result there may be a threat to plant safety.
This event will be escalated based upon system malfunctions or damage consequences.
REFERENCE(S)
- 1. 2000-ABN-3200.30 "Control Room Evacuation" NUMARC IC HS2 DIFFERENCES None Page 103 of 122 Revision 0c
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Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HA2 INITIATING CONDITION Control Room Evacuation Initiated EAL THRESHOLD VALUES Entry into 2000-4kBN-3200.30 "Control Room Evacuation" MODE APPLICABILITY:
All BASIS:
Control Room evacuation requires establishment of plant control from outside the control room (e.g., local control and remote shutdown panel) and support from the Technical Support Center and/or other emergency facilities as necessary. Control Room evacuation represents a serious plant situation since the level of control is not as complete as it would be without evacuation. The establishment of system control outside of the Control Room will bypass many protective trips and interlocks. In addition, many of the instruments and assessment tools available in the Control Room will not be available.
This event will be escalated to a Site Area Emergency if control cannot be established within fifteen minutes.
REFERENCE(S)
- DIFFERENCES None Page 104 of 122 Revision Oc
Ovster O- str Creek Creek..
Nuclear Station Annex Nu... ear.. tati..n.Ann....................
Exelnn Nnel*ar Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HA3
'INITIATING CONDITION Natural OR Destructive Phenomena Affecting a Vital Area EAL THRESHOLD VALUE
- 1. Confirmed earthquake requiring reactor scram in accordance with 2000-ABN 3200.38 Station Seismic Event OR
- 2. Tornado or sustained wind speeds > 100 mph causing damage to Plant Vital Structures (Table H-1)
- 3. Report of visible structural damage to ANY Plant Vital Structure (Table H-i) due to natural or destructive phenomena OR
- 4. Vehicle crash damaging or affecting Plant Vital Structure (Table H-i)
- 5. Abnormal Intake Structure level, as indicated by EITHER:
- > 6.0 ft. MSL (> 4.92 psig on PI-SWS-1 [2])
- <-4.0 ft. MSL (<0.50 psig on PI-533-1172 or PI-533-1173)
MSL = Mean Sea Level Table H-1 Plant Vital Structures Reactor Bldg.
Turbine Bldg.
Control Room Complex Main Transformer/Condensate Transfer Pad Intake Structure
- 1 EDG Vault
flvefpr ProoUL N"AarI~ faftnn A nnixv 1W Ian ATX.nlI Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HA3 - Cont'd BASIS These threshold values are natural or destructive phenomena, which represent actual or potential substantial degradation of the level of safety of the plant. The affects of the phenomena should also be evaluated on a system or component basis in relation to Technical Specification and evaluated for further classification via the System Malfunction and Fission Product Barrier Recognition Categories. The EAL thresholds associated with this IC escalate from the Unusual Event EALs in HU3 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by control indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial "report" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation. Escalation to higher classifications occur on the basis of other initiating conditions (e.g., System Malfunction).
Threshold Value 1 - This EAL addresses a confirmed earthquake that affects safe plant operation by jeopardizing the availability of safety systems, systems required to complete safe shutdown, or causing spurious actuation of equipment and warranting a manual reactor scram. A call to the Lamont-Doherty Geological Observatory is the primary confirmation source. Other confirmation includes reports from television or radio stations, or reports from university monitoring stations. An earthquake of this magnitude may be sufficient to cause damage to safety related systems and functions. This EAL threshold is intended to be consistent with the Operating Basis Earthquake (OBE) for OCNS which is 0.11 g per FSAR Update Section 3.7. Confirmation of the magnitude of a seismic event may be confirmed with Lamont-Doherty Geological Observatory.
Threshold Value 2 - This EAL is based on the 100 year storm (0-50 ft.) per FSAR Update Section 3.3.1. Wind loads of this magnitude can cause damage to safety functions. This EAL addresses events where Plant Vital Structures have been struck with high winds, and thus damage may have occurred to safe shutdown systems. No attempt should be made to assess the magnitude of damage to Plant Vital Structures prior to classification.
Threshold Value 3 - The threshold value of this EAL should be determined relative to the damage that might occur from events described in Threshold Values 1 & 2. This EAL specifies the Plant Vital Structures, which contain systems and functions required for safe shutdown of the plant.
Page 106 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HA3 - Cont'd BASIS - Cont'd Threshold Value 4 - This criteria address crashes of vehicles that have caused damage to Plant Vital Structures, and thus damage may be assumed to have occurred to safe shutdown systems. No attempt should be made to assess the magnitude of damage to Plant Vital Structures prior to classification. The evidence of damage is sufficient for declaration. A vehicle crash includes aircraft and large motor vehicles, such as a crane.
Threshold Value 5 - High Intake Structure level, > 6.0 feet above MSL (> 4.92 psig on PI-SWS-1 [2]) is capable of causing flooding that can affect Plant Vital Structures. At levels > 6.5 ft. above MSL, Circulating Water Pumps may become flooded. At levels > 8.0 ft. above MSL, Service Water pumps may become flooded. No attempt should be made to determine the magnitude of flooding. This is a long lead time event but this level is at the intake structure lower deck so classification as an Alert Event is appropriate. The evidence of flooding is sufficient for declaration.
Low Intake Structure level <-4.0 feet above MSL (< 0.50 psig on PI-533-1172 or PI 533-1173) indicates the possible loss of Emergency Service Water pumps. Procedures require the unit to be brought to cold shutdown.
This event will be escalated to a higher emergency classification based upon damage consequences covered under other various EAL Sections.
REFERENCE(S)
- 1. 2000-ABN-3200.38 Station Seismic Event
- 2. FSAR Update Section 3.3.7 (Seismic)
- 3. FSAR Update Section 3.3.1 (High winds)
- 4. 2000-ABN-3200.31 High Winds
- 5. 2000-ABN-3200.32 Response to Low Intake Levels
- 6. 2000-ABN-3200.29 Response to Fire (Plant Vital Structures)
ýNUMARC iC HA1 DIFFERENCES
- 1. OCNS does not have installed seismic instrumentation to determine if seismic activity is in excess of OBE levels. Procedure 2000-ABN-3200.38 "Station Seismic Event" requires the Shift Manager to scram the reactor for conditions in which the seismic activity causes a threat to safe plant operation. This is consistent with earthquakes in excess of OBE levels.
- 2. NUMARC IC HA1.6 - No safety related plant areas are susceptible to turbine failure generated missiles.
Page 107 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HU3 INITIATING CONDITION Natural OR Destructive Phenomena Affecting the Protected Area
- EAL THRESHOLD VALUE
- 1. Felt earthquake OR
- 2. Report by plant personnel of a tornado strike within the Protected Area OR
- 3. Sustained wind speeds > 75 mph as indicated by on-site meteorological instrumentation OR
- 4. Vehicle crash within the Protected Area Boundary that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant.
- 5. Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.
- 6. Abnormal Intake Structure level, as indicated by EITHER:
"* > 4.5 ft. MSL (>4.26 psig on PI-SWS-1 [2])
"* <-3.0 ft. MSL (<0.94 psig on PI-533-1172 or PI-533-1173)
MSL = Mean Sea Level MODE APPLICABILITY:
ALL BASIS:
These EALs are categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators. Escalation of the event to an Alert occurs when the magnitude of the event is sufficient to result in damage to equipment contained in the specified location.
Threshold Value 1 - This EAL addresses a felt earthquake. A felt earthquake is considered to be an earthquake of sufficient intensity such that the vibratory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of control room operators on duty at the time. An earthquake of this magnitude may be sufficient to cause minor damage to plant structures or equipment within the Protected Area. Damage is considered to be minor, as it would not affect physical or structural integrity. This event is not expected to affect the capabilities of plant safety functions. Tfiis event will be escalated to an Alert if the earthquake reaches thle level requiring a reactor scram per 2000-ABN-3200.38 Station Seismic Event.
Page 108 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HU3 - Cont'd
,BASIS: - Cont'd Threshold Values 2 & 3 - A tornado touching down within the Protected Area or sustained wind speeds > 75 mph within the Owner Controlled Area are of sufficient velocity to have the potential to cause damage to Plant Vital Structures. The value of 75 mph was selected to coincide with the Beaufort Scale for Hurricane wind speed winds of 73-136 mph. These criteria are indicative of unstable weather conditions and represent a potential degradation in the level of safety of the plant. Verification of a tornado will be by direct observation and reporting by station personnel. Verification of sustained (> 15 minutes in duration) wind speeds > 75 mph will be via meteorological data in the control room. This event will be escalated to an Alert if the tornado or high wind speeds result in damage to Plant Vital Structures.
Threshold Value 4 - This criterion is intended to address such items as plane, helicopter, or train crash that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant. If the crash is confirmed to affect a plant vital area structure, the event may be escalated to an Alert classification.
Threshold Value 5 - This criterion is intended to address main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustible fluids (e.g., lubricating oils) and gases (e.g., hydrogen) to the plant environs.
Actual fires and flammable gas build up are appropriately classified via other EALs.
Turbine failure of sufficient magnitude to cause observable damage to the turbine casing or seals of the turbine generator raises the potential for leakage of combustible fluids and gases (Hydrogen cooling) to the Turbine Building. The damage should be readily observable and should not require equipment disassembly to locate.
Threshold Value 6 - High Intake Structure level, > 4.5 feet above MSL (> 4.26 psig on PI-SWS-1 [2]) is sufficiently high to require plant shutdown per 2000-ABN-3200.31 High Winds. This event will be escalated to an Alert classification based on water level reaching the elevation of the Intake Structure lower deck.
Low Intake Structure level <-3.0 feet above MSL (< 0.94 psig on PI-533-1172 or PI 533-1173) indicates the possible loss of Radwaste Service Water pumps and approaching levels which may result in a loss of vital cooling equipment. This event will be escalated to an Alert based upon water level dropping to < -4.0 feet above MSL.
Page 109 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HU3 - Cont'd REFERENCE(S)
- 1. 2000-ABN-3200.38 Station Seismic Event
- 2. FSAR Update Section 3.3.7 (Seismic)
- 3. FSAR Update Section 3.3.1 (High winds)
- 4. 2000-ABN-3200.31 High Winds
- 5. 2000-ABN-3200.32 Response to Low Intake Levels
Page 110 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Exelon Nuclear Nuclear Oyster Creek Nuclear Station Annex II Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HA4 INITIATING CONDITION Fire OR Explosion Affecting Operability of Safety Systems Required for Safe Shutdown EAL THRESHOLD VALUES Fire or explosion causing damage to a Plant Vital Structure (Table H-l) or affecting one or more Safe Shutdown Systems (Table H-2)
AND Safe Shutdown System operability is required Table H-1 Plant Vital Structures Reactor Bldg.
Turbine Bldg.
Control Room Complex Main Transformer/Condensate Transfer Pad Intake Structure
- 1 EDG Vault
- 2 EDG Vault EDG Fuel Oil Storage Tank Table H-2: Safe Shutdown Systems
- 4160 Safeguard Busses
- Containment Spray 0 Condensate Transfer
- Service Water MODE APPLICABILITY:
ALL BASIS:
Explosion - A rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to nearby structures or equipment.
Page 111 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HA4 - Cont'd
'BASIS: - Cont'd Fire - 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 constitute fires. Observation of a flame is preferred, but is NOT required if large quantities of smoke and heat are observed.
The primary concern of this EAL is the magnitude of the fire or explosion 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. 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, potential inoperability may be determined by visual observation and other control room indications such as loss of indicating lights.
2000-ABN-3200.30 Control Room Evacuation was consulted to determine systems included in Table H-2 Safe Shutdown Systems.
In those cases where it is believed that the fire may have caused damage to Safety Shutdown 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 Shutdown 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 actual magnitude of damage. The occurrence of the fire or explosion with reports of damage (e.g., deformation, scorching) is sufficient for declaration.
REFERENCE(S)
- 1. 2000-ABN-3200.29 Response to Fires
- 2. 2000-ABN-3200.30 Control Room Evacuation NUMARC IC HA2 DIFFERENCES None Page 112 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HU4 INITIATING CONDITION Fire Within Protected Area Boundary NOT Extinguished in < 15 min. of Detection
'EAL THRESHOLD VALUE
- 1. Fire within or contiguous to a Plant Vital Structure (Table H-i)
AND Fire is NOT extinguished in < 15 min. of EITHER:
"* Control Room notification OR
"* Verification of alarm OR
- 2. Report by plant personnel of an unanticipated explosion within the Protected Area Boundary resulting in visible damage to a Plant Vital Structures (Table H-i)
Table H-1 Plant Vital Structures Reactor Bldg.
Turbine Bldg.
Control Room Complex Main Transformer/Condensate Transfer Pad Intake Structure
- 1 EDG Vault
- 2 EDG Vault EDG Fuel Oil Storage Tank MODE APPLICABILITY ALL BASIS Verification - Determination is made that the fire alarm is not spurious.
Explosion - A rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to nearby structures or equipment.
Page 113 of 122 Revision Oc
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HU4 - Cont'd BASIS - Cont'd The purpose of this EAL is to address the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems. This excludes such items as fires within administration buildings, wastebasket fires, and other small fires of no safety consequence. This EAL applies to buildings and areas contiguous to plant vital areas or other significant buildings or areas. The intent of this IC is not to include buildings (e.g.,
warehouses) or areas that are not contiguous or immediately adjacent to plant vital areas.
This EAL addresses fires in Plant Vital Structures that house Safe Shutdown Systems.
These fires may be precursors to damage to safety systems contained in these structures.
Verification that a fire exists is by operator actions to confirm that fire alarms received in the Control Room are not spurious or by any verbal notification by plant personnel. Fifteen minutes has been established to allow plant staff to respond and control small fires or to verify that no fire exists.
This event will be escalated to an Alert if the fire damages Safe Shutdown Systems required for the current operating condition.
REFERENCE(S)
- 1. Unanticipated explosions from NUMARC IC HU1.5 has been incorporated into this EAL as a logical precursor to OCNS EAL HA4.
Page 114 of 122 Revision Oc
Oyster Creek Nuclea.r qtatinn Anney Exelan Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HA5 INITIATING CONDITION Release of Toxic OR Flammable Gases Within a Facility Structure Which :
Jeopardizes Operation of Systems Required to Maintain Safe Operation OR to Establish or Maintain Cold Shutdown
,EAL THRESHOLD VALUE
- 1. Report or detection of toxic gases within Plant Vital Structures (Table H-1) in concentrations that will be life threatening to plant personnel.
- 2. Report or detection of flammable gases within Plant Vital Structures (Table H-1) in concentrations affecting the safe operation of the plant Table H-1 Plant Vital Structures Reactor Bldg.
Turbine Bldg.
Control Room Complex Main Transformer/Condensate Transfer Pad Intake Structure
- 1 EDG Vault
- 2 EDG Vault EDG Fuel Oil Storage Tank MODE APPLICABILITY ALL JBASIS This IC is based on gases that affect the safe operation of the plant. This IC applies to buildings and areas contiguous to Plant Vital Structures. The intent of this IC is not to include buildings (e.g., warehouses) or other areas that are not contiguous or immediately adjacent to Plant Vital Structures. It is appropriate that increased monitoring be done to ascertain whether consequential damage has occurred. Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Levels / Radioactive Effluent, or Emergency Director Judgment ICs.
Page 115 of 122 Revision 0c
Ovster Creek Nuclear Station Annex Exelon Nuclear Oyster Creek...N..cl......St t...n..Exelor Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HA5 - Cont'd BASIS - Cont'd EAL #1 is met if measurement of toxic gas concentration results in an atmosphere that is Immediately Dangerous to Life and Health (IDLH) within a Plant Vital Structures or any area or building contiguous to Plant Vital Structures. Exposure to an IDLH atmosphere will result in immediate harm to unprotected personnel, and would preclude access to any such affected areas. Areas that require only temporary access that can be supported by the use of respiratory protection should not be considered as exceeding this threshold.
EAL #2 is met when the flammable gas concentration in a Plant Vital Structure or any building or area contiguous to a Plant Vital Structure exceed the Lower Flammability Limit. Flammable gasses, such as hydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repair equipment/components (acetylene - used in welding). This EAL addresses concentrations at which gases can ignite/support combustion. An uncontrolled release of flammable gasses within a facility structure has the potential to affect safe operation of the plant by limiting either operator or equipment operations due to the potential for ignition and resulting equipment damage/personnel injury. Once it has been determined that an uncontrolled release is occurring, then sampling must be done to determine if the concentration of the released gas is within this range.
REFERENCE(S)
Page 116 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HU5 INITIATING CONDITION Release of Toxic OR Flammable Gases Deemed Detrimental to Safe Operation of the Plant.
,EAL THRESHOLD VALUES
- 1. Report or detection of toxic or flammable gases that could enter the site boundary in amounts that can affect normal operation of the plant OR
- 2. Report by Local, County or State officials for potential evacuation of site personnel based on an offsite event
ýMODE APPLICABILITY ALL
'BASIS:
This IC is based on the existence of uncontrolled releases of toxic or flammable gas that may enter the site boundary and affect normal plant operations. It is intended that releases of toxic or flammable gases are of sufficient quantity, and the release point of such gases is such that normal plant operations would be affected. This would preclude small or incidental releases, or releases that do not impact structures needed for plant operation.
The EALs are intended to not require significant assessment or quantification. The IC assumes an uncontrolled process that has the potential to affect plant operations, or personnel safety.
A gas release is considered to be impeding normal plant operations if concentrations are high enough to restrict normal operator movements. It also includes areas where access is only possible with respiratory equipment, as this equipment restricts normal visibility and mobility. It should not be construed to include confined spaces that must be ventilated prior to entry or situation involving the Fire Brigade who are using respiratory equipment during the performance of their duties unless it also affects personnel not involved with the Fire Brigade.
An offsite event (such as a tanker truck accident or train derailment releasing toxic gases) may place the Protected Area within the evacuation area. This evacuation is determined from the DOT Evacuation Tables for Selected Hazardous Materials in the North American Response Guidebook for Hazardous Materials.
Escalation of this EAL is via HA5, which involves a quantified release of toxic or flammable gas affecting Plant Vital Structures.
Page 117 of 122 Revision 0c
lV~dlrn VTwAnlat Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HU5 - Cont'd REFERENCE(S)
- 1. 2000-ABN-3200.33 Toxic Materials/Flammable Gas Release NUMARC IC HU3 DIFFERENCES None Page 118 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HG6 INITIATING CONDITION' Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency EAL THRESHOLD VALUES
- 1. Actual or imminent core degradation with potential loss of containment.
- 2. Potential uncontrolled radionuclide release, which can reasonably be expected to exceed 1 Rem TEDE, or 5 Rem CDE Child Thyroid plume exposure levels at the Site Boundary or beyond
,MODE APPLICABILITY ALL BASIS General Emergency - Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Imminent - Mitigation actions have been ineffective and trended information indicates that the event or condition will occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Potential - Mitigation actions are not effective and trended information indicates that the parameters are outside desirable bands and not stable or improving.
This EAL allows the Emergency Director to declare a General Emergency upon the determination of an actual or imminent substantial core degradation or melting with the potential for loss of containment integrity, but is not explicitly addressed by other EALs.
Releases may exceed the EPA Protective Action Guidelines for more than the immediate site area and will be classified under Event Category R, "Abnormal Radiological Levels/Effluents".
REFERENCE(S)
None NUMARC IC HG2 DIFFERENCES None Page 119 of 122 Revision 0c
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Exel-n Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HS6 INITIATING CONDITION Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of Site Area Emergency EAL THRESHOLD VALUE Other conditions exist which in the judgment of the Emergency Director indicate actual or likely major failures of plant functions needed for protection of the public.
MODE APPLICABILITY ALL BASIS Site Area Emergency - Events are in process or have occurred which involve actual or likely failure of plant functions needed for protection of the public. Any releases are not expected to result in exposure levels, which exceed EPA Protective Action Guideline exposure levels except near the site boundary.
This EAL allows the Emergency Director to declare a Site Area Emergency upon the determination of an actual or likely major failure of plant functions needed for protection of the public, but is not explicitly addressed by other EALs.
Releases are not expected to result in exposure levels, which exceed the EPA Protective Action Guidelines except within the site boundary and will be classified under Event Category R, "Abnormal Radiological Levels/Effluents".
REFERENCE(S)
None NUMARC IC HS3 DIFFERENCES None Page 120 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND *OTHER CONDITIONS HA6 INITIATING CONDITION Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert
,EAL THRESHOLD VALUE Other conditions, exist which in the judgment of the Emergency Director indicate that plant safety systems may be degraded and that increased monitoring of plant functions is warranted
,MODE APPLICABILITY ALL
,BASIS Alert - Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
This EAL allows the Emergency Director to declare an Alert upon the determination that the level of safety of the plant has substantially degraded but is not explicitly addressed by other EALs. This includes a determination by Shift Management that the TSC and OSC should be activated and command and control functions should be transferred for the event to be effectively mitigated. Transfer of command and control functions is used as an initiator since an event significant to warrant transfer is a substantial reduction in the level of safety of the plant.
REFERENCE(S)
None NUMARC IC HA6 DIFFERENCES None Page 121 of 122 Revision 0c
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HU6 INITIATING CONDITION Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Unusual Event IEAL THRESHOLD VALUE Other conditions exist which in the judgment of the Emergency Director indicate a potential degradation in the level of safety of the plant MoDE APPLICABILITY ALL BASIS (References)
Unusual Event - Events are in process of have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
This EAL allows the Emergency Director to declare an Unusual Event upon the determination that the level of safety of the plant has degraded. Where the degradation is associated with equipment or system malfunctions, the decision that it is degraded should be made upon functionality, not operability. A system, subsystem, train, component or device, though degraded in equipment condition or configuration, should be considered functional if it is capable of maintaining respective system parameters within acceptable design limits.
Releases of radioactive materials requiring offsite response or monitoring are not expected to occur at this level unless further degradation of safety systems occurs. However, if one does occur, it will be classified under Event Category R, "Abnormal Radiological Levels/Effluents" REFERENCE(S)
None NUMARC IC HU5 DIFFERENCES None Page 122 of 122 Revision Oc
Ovwter Creek Nnulear Station Annex
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SI IV Exelon Nuclear g
m TABLE OCNS 3-1: Emergency Action Level (EAL) Matrix GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ABNORMAL RAD LEVELS / EFFLUENTS RGI Actual or Projected Site Boundary Dose 2 3 RS1 Actual or Projected Site Boundary Dose 2 3RAI Release>200XODCMLimtfor R Using Actual Meteorology: Using Actual Meteorology: > 15 min. LLRUI Relea>s60 m>.2 O ML tf
> 1000 mRem TEDE > 100 mRem TEDE EAL Threshold Value- EAL Threshold Value OR OR I1. Unplanned radiological release lasting >15 min. in excess of 1. Unplanned radiological release lasting> 60 mm. inmexcess of
> 5000 mRem CDE Child Thyroid > 500 mRem CDE Child Thyroid Table RI "Alert" thresholds Table RI "Unusual Event" threshold
- E EAL Threshold Value: , EAL Threshold Value: AND i AND I 1. Radiological release in excess of Table RI "General 1. Radiological release in excess of Table RI "Site Area Releases CANNOT be determined in < 15 min. (from time Releases CANNOT be determined in <560 mi. (from time Emergency" threshold Emergency" threshold Table RI threshold was exceeded) to be below Table R2 Table RI threshold was exceeded) to be below Table R2
""AND AND "Alert" thresholds "Unusual Event" thresholds Releases CANNOT be determined in <515 min. (from time Releases CANNOT be determined in < 15 min. (from time OR oR
- 6 Table RI threshold was exceeded) to be below Table R2 Table RI threshold was exceeded) to be below Table R2 "Site 2. Unplanned radiological releases lasting > 15 min. in excess of: 2. Unplanned radiological releases lasting > 60 min. in excess of C4Dose Assessment "General Emergency" thresholds Area Emergency" thresholds ANY Table R2 "Alert" threshold ANY Table R2 "Unusual Event" threshold OR OR
- 2. Radiological releases exceed ANY Table R2 "General :2. Radiological releases exceed ANY Table R2 "Site Area Emergency" threshold Emergency" threshold
.RA2 In-Plant Radiation Levels Impede Plant RU2 Rise In Plant Radiation Levels by a EAL
- Operations Thresholdreadings
[EilP Value- > 15 mR/hr in EITHER of the following.
.i : Factor of 1000 EF JJ
- 1. Radiation EAL Threshold Value C!C" Radiaonadin RmRnhro m EITHEn 1. Valid area radiation monitor readings indicate an unplanned *
.MinOnrR oo: rise by a factor of 1000 over normal levels as detected by None None Central Alarm Station: either permanent or temporarily installed radiation monitors or' OR by manual survey
- 2. In-plant radiation readings > 1 R/hr in areas requiring access RU3 High Off-gas [ f Sinorder to maintain safe operatin or performn a safe " QA Threshold Value Radiation Levels
- 1. Off-gas radiation reading > HI-Hi alarm value for > 15 min.
shutdown
- RU4 High coolant activity T.'S, .
- None
- None NoneEAL Threshold Value, 0 gE N. 1 Reactor coolant activity > 0.2 RCi/gmDEI TEhTableR2 Dose Assessment Thresholds Table RI - Effluent Monitor Thresholds
- At or beyond the Site Boundary based on a I hour release duration Release Point/Monitor General Emergency Site Area Emergency Alert Unusual Event Method General Emergency Site Area Emergency Alert Unusual Event Main Stack RAGEMS Sample >200X ODCM4 6 1.14 >2XODCM46 1.14 Torus/Drywell2" Vent >21.0 RCm/ccHRM >21 PCi/cceHRM >CPSA > CPS U Analysis N/A N/A OR OR rsrwe> 200 XODCM 46.1.1.5 > 2X ODCM 4 611 via SBGT Reactor Bldg. via SBGT >3 0 itCi/cc HRM > 0 3 pCVdcc IRM > CPS A > CPS U Field Team > 1000 mRem/hr Whole Body >100 mRem/hr Whole Body > 10 mRem/hr Whole Body Monitoring__OR OR OR Not Applicable Turbine Bldg. RAGEMS ontoring* >5000 mRem CDE Child Thyroid >500 mRem CDE Child Thyroid > 34 mRem CDE Child Thyroid Via EFI-4 & EFI-33 > 3 0 pCi/cc fIRM > 0.3 VCi/cc fIRM >200 X tli-Hi > 2 X Hi-Hi alarm setpoint alarm serpoint Dose > 1000 mRem TEDE > 100mRem TEDE > 10 mRem/hrTEDE >0 10mRem/hr TEDE Assessment* OR OR OR OR Service Water Effluent N/A N/A >3.13 E4 cpm >415 E2 cpm >5000 mRem CDE Child Thyroid > 500 mRem CDE Child Thyroid >*34 mRem CDE Child Thyroid > 0 34 mRem CDE Child Thyroid Plant Modes: *--* Power Operations W Hot Shutdown (>- 212 IF) j 'j Cold Shutdown (< 212 IF) VE Refuel,,E Defueled HRM= High Range Monitor LRM= Low Range Monitor Page 7 of 122 Revision 0e
Oyster Creek Nuclear Station Annex TA V' -.... .I ............. ...... ....... . Exelon Nuclear IAULrn NSJ-1:
3n; alli *lttny AtlonL evel (I L) MiVrtxl(loL0"1 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ABNORMAL RAD LEVELS / EFFLUENTS RA5 Major Damage OR Uncovering of 12 3 4 RU5 Potential Damage OR Uncovering of Spent Spent Fuel. Fuel EAL Threshold Value EAL Threshold Value:
None None :1. Valid unanticipated Hi alarm on one or more Refuel Floor 1. Uncontrolled water level drop in the Spent Fuel Pool with ARMs (Table R-3) ALL irradiated fuel assemblies remaining covered by water OR
- 2. Report of visual observation that irradiated fuel in the Spent
_ _ _ _ _ __ : Fuel Pool or Fuel Transfer Canal is uncovered RA6 Loss of Water Level That Has OR Will r - RU6 Uncontrolled Water Level Drop in Uncover Irradiated Fuel in the Reactor Reactor Cavity "Cavity EAL Threshold Value:
None None EAL Threshold Value: 1. Unexpected Skimmer Surge Tank Lo-Lo level alarm
- 1. Report of visual observation that irradiated AND fuel in the Reactor Cavity is or will be Visual observation of an uncontrolled drop in uncovered water level below the fuel pool skimmer surge >
tank inlet
, RU7 Independent Spent Fuel Storage Installation [ 7f" I-S1. Threshold Ti:EAL Value:
Radiation readings > 10 times normal at ANY of the following ISFSI locations:
None None None'
- on contact with roof
- on contact with shield wall Table R-3: Refuel Floor ARMs
- C-5, Cnt Mon
- C-9, North Wall
- C-10, North Wall 0 B-9, Open Floor Page 8 of 122 Revision Oc
vsefr C(reeppk Nu.lper Satinn Anney Ir uysr
........... ................. Exelon Nuclear TABLE UOINS 3-1: Emergency Action Level (EAL) Matrix (Cont'd)
FISSION PRODUCT BARRIER MATRIX (Applicability: Modes 1 & 2 ONLY) [1_12111 _
FISSION PRODUCT BARRIER STATUS FGI: GENERAL EMERGENCY FSI: SITE AREA EMERGENCY FAI: ALERT FUll: UNUSUAL EVENT Fuel Clad - LOSS X X X X X X Fuel Clad - POTENTIAL LOSS X X X X X Reactor Coolant System - LOSS X X X X XXX Reactor Coolant System - POTENTIAL LOSS X X X Primary Containment - LOSS X X X X X X Primary Containment - POTENTIAL LOSS XI-_X
- 1. FUEL CLAD BARRIER 2. REACTOR COOLANT SYSTEM BARRIER 3. PRIMARY CONTAINMENT BARRIER LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS
- a. RPV Water Level OR OR lowered by procedure) None NoneNos None I. Entry into SAMGs as required by OR E~
CANNOT be determined CANNOT be determined CANNOT be determined
- b. Drywell Radiation I. Containment Hi-Range Radiation 1. Containment Hi-Range Radiation I. Containment Hi-Range Radiation Monitoring Monitoring
> 440 fl/hr System (CHRRMS) None Monitoring
>45 fl/hr System (CHRRMS) None None Monitoring System (CHRRMS)
> 2.0E+4 fl/hr
- 1. Rapid, unexplained drop in drywell 1.Diywell pressure >3.0 Ig pressure following an initial rise
- c. Drywell Pressure None None AND None OR Indication of RCS leak inside Drywell 2. Drywell pressure response not consistent 3. Drywell pressure > 44 psig with LOCA conditions indicating a containment breach I Coolant activity> 300 pCi/gm (DEl)noutside 1. Unisolable Main Steam Line break 3. RCS Leakage > 50 gpm 1. Failure of all isolation valves in ANY containment one line penetrating Primary OR Containment to close when required OR 4 Unisolable primary system leakage AND
- 2. Unisolable Isolation Condenser tube outside of drywell as indicated by Downstream pathway exists to rupture exceeding EITHER of the following in environment one or more areas requiring a scram- OR
"* EMG.3200 11 Max Normal 2. Intentional venting per EMG-3200 02 is Temperature required with Drywell pressure> 3.0
- d. Breached / Bypassed None
"* EMG-3200 OR 11 Max Normal psig OR None Radiation Level 3 Unisolable primary system leakage outside of drywell as indicated by exceeding EITHER of the following in one or more areas requiring a scram:
"* EMG-3200.11 Max Normal Temperature OR
"* EMG-3200 11 Max Normal Radiation Level
- e. Containment I Containment 112concentration > 6 %
Hydrogen None None None None None AND Concentration Containment 02 concentration > 5 %
- f. Emergency Director 1. ANY condition in the judgment of the Emergency Director that indicates Loss or 1. ANY condition in the judgment of the Emergency Director that indicates Loss or 1. ANY condition in the judgment of the Emergency Director that indicates Loss Judgment Potential Loss of the Fuel Clad barrier. Potential Loss of the RCS barrier or Potential Loss of the Primary Containment barrier Page 9 of 122 Revision Oc
flqtpr rFran~kN1%3.q.siantinfln A nnpy 9.uvsyer %.reeKiNuciear marion Annex Exelon Nuclear TABLE OCNS 3-1: Emeriency Action Level (EAL) Matrix (Cont'd)
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTIONS F'
MGI Prolonged Loss of ALL Offsite AC 1 MSI Loss of ALL Offsite AC Power AND MAI AC power capability to essential busses ss of All Offsite Power to Essential Power AND Prolonged Loss of ALL Onsite AC Power 1 .. Loss of ALL Onsite AC Power to EsnilBse reduced to a single power source for greater than 15 Busses for Greater Than 15 Mm. IIIIIh
- OnsiTesACholer A EssentialBues minutessuch that any additional single failure would result EAL Threshold Value:
EAL Threshold Value- EAL Threshold Value: I osoofoffsite fst power oe tooBT 10 Busses ussi1C and n
- 1. BOTH 4160VBussesIC and ID de-energized for> 15min. :1. BOTH 4160V BussesIC andIlDde-energizedfor> 15 mm. in station blackout 1. Loss BOTH 4160V I D for > 15 min.
AND EAL Threshold ValueD ANY of the following: 1. Loss ofoffsite power to BOTH 4160V Busses IC and 1D for t
- Restoration ofat least one emergency bus within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> > 15mrin Sis not likely AND a RPV level CANNOT be maintained > 0" TAF OR EITHER of the 4160V busses IC or ID de-energized for CANNOT be determined *".... > 15 min
".................. °...............o...................°.................
Torus water temperature and RPV pressure exceeds the M Heat Capacity Temperature Limit (Figure F, EMG- :MA2 Loss of All Offsite Power AND Loss of All 3200.02) Onsite AC Power to Essential Busses During Cold Shutdown Or Refueling Mode EAL Threshold Value:
- 1. BOTH 4160V Busses 1C and ID de-energized for > 15 min.
MS3 ~~~EAL Loss of All VitalThreshold Value DC Power [ MU3 Unplanned Loss of Required DC ET _M A Te l a:i-iPower During Cold Shutdown or L .
Non Loss of ALL vital DC power indicated by < 115 VDC
.None NoneN Refueling Mode> 15 Mi .
ReulnMoe>!5M indication on 125 VDC Busses B and C for > 15 main. EAL Threshold Value
- 1. Loss of ALL vital DC power indicated by < 115 VDC indication on 125 VDC Busses B and C for > 15 min.
MG4 Auto and Manual SCRAM NOT Successful, AND Loss of Core Cooling or IT I fl : MS4 Auto and Manual SCRAM NOT Successful Ftl ] MA4 Auto SCRAM NOT Successful Valu IEI))
Heat Sink EAL Threshold Value: EITHER EAL Threshold Value i1. RPS setpoint for an automatic SCRAM exceeded 2 :1. RPS setpoint for an automatic SCRAM exceeded AND :1. RPS setpoint far an automatic SCRAM exceeded "AND Failure of automatic RPS, ARI and manual SCRAM to AND ii Failure reactor reduce of automatic < 2% ARI and manual SCRAM to power RPS, reduce reactor power <2% :O Failure of automatic SCRAM to achieve reactor shutdown None AND EITHER: :2. Loss of manual SCRAM capability indicated by failure of ALL
- RPV level CANNOT be restored and maintained manual SCRAM attempts to achieve reactor shutdown
> -30" TAF OR CANNOT be determined OR
- Torus water temperature and RPV pressure exceeds the Heat Capacity Temperature Limit (Figure F, EMG-3200 02)
MS5 Complete Loss of Functions Needed to MA5 Inability to Maintain Plant in Cold Shutdown Achieve AND Maintain Hot Shutdown f EAL Threshold Value:
SEAL Threshold Value: 1. Unplanned loss of all Technical Specification required systems
- 1. Torus water temperature and RPV pressure CANNOT be available to provide decay heat removal functions Nmaintained below the Heat Capacity Temperature Limit AND None(Figure F, EMG-3 200 02) Uncontrolled temperature rise that approaches or exceeds 212 OF None Ci Page 10 of 122 Revision 0e
Oyster Creek Nuclear Station Annex Exelon Nuclear Ir r7An,*T1-V ,- rWQ
- 1[- .1,-l6":-- A* J* T - d 1(VAI t * .&# ,!- r
'FABLE uJLS .3-i:EmergencV ACtiOn Level (EAL) Matrix ( font'd)
GENERAL EMERGENCY 7 7 SITE AREA EMERGENCY ALERT UNUSUAL EVENT SYSTEM MALFUNCTIONS (cont.)
WMS6 Inability to Monitor a Significant Transient 1 -'MA6Unplanned Loss of Most or All Safety System IMU6 Unplanned Loss of Most or All Safety System In Progress i Annunciation or Indication in Control Room Annunciation or Indication in the Control E M T EAL Threshold Value- With Either (1) a Significant Transient in Progress, or (2) Room > 15 Min.
- 1. A significant transient is in progress (Table M-1) Compensatory Non-Alarming Indicators are Unavailable EAL Threshold Value
- AND folwnarlst EAL Threshold Value: :1. Unplanned loss, for> 15 min, of MOST (Note 1) or all of
- ALL of the following are lost: 1. Unplanned loss, for > 15 min. of MOST (Note 1) or all of EITHER:
Safety system annunciators (Table M-2) EITHER:
- Safety system annunciators (Table M-2)
None
- Safety function indicators (Table M-3) a Safety system annunciators (Table M-2) OR
,
- Plant Process Computer OR
- Safety function indicators (Table M-3) 0 Safety function indicators (Table M-3)
AND EITHER:
- A significant plant transient isin progress (Table M-1)
- Plant Process Computer isunavailable MS7 Loss of Water Level in the Reactor Vessel MU7 RCS Leakage That Has or Will Uncover Fuel in the I 47 EAL Threshold Value: a None Reactor Vessel None 1. Unidentified leakage> 10 gpm EALThreshold Value: OR
- 1. RPV level <0 "TAF OR CANNOT be determined 2. Identified leakage > 25 gpm MU8 Unplanned Loss of ALL Onsite OR "0 Offsite Communications Capabilities C1 EAL Threshold Value:at None None None 1. Loss of all onsite communications (Table M-4) affecting the .
Eability to perform routine operations OR "
- 2. Loss of all offsite communications (Table M-4)
MU9 Plant is not brought to required operating
- mode within Technical Specifications LCOFL3ML None Action Statement Time .
None None EAL Threshold Value:
- 1. Required operating mode is NOT reached within Tech Spec.
LCO action completion time Table M-1: Significant Plant Transients Table M-2: Safety System Annunciators Table M-3: Safety Function Indicators Table M4: Communications NOTE I
- Scram "*ECCS (B, C) "* Reactor Power, Pressure and Level Onsite Offsite Offsite "MOST" refers to a loss of~-75% or a significant
- > 25% thermal power change "* Containment Isolation (G,11,J) (Panel 4F. 5F, 6F) "*Plant Paging
- Conventional tel. lines & ED Hotline risk that a degraded plant condition could go
- Sustained power oscillations "*Reactor Scram (G) "*Decay Heat Removal (Panel IF/2F) System
- Cell Phones
- New Jersey State Police (NJSP undetected. Use is not intended to require a (30 watts/cm 2 LPRM peak to peak) "* Process Radiation Monitormg (IOF) "*Conventional
- Radio Notification Line) detailed count of annunciators/indicators.
0 Stuck Open EMRVs
"*Containment Safety Functions telephone lines
- NJ State ED Hotline
"*Radio HPN
- Environmental Assessment Direct 9 Bureau of Nuclear Engineenng Line Information Line Page I I of 122 Revision Oc
Oyster Creek Nuclear Station Annex Exlo.*n Nulear*n, TABLE OCNS 3-1: Emergency Action Level (EAL) Matrix (Cont'd)
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT 7 UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS HGI SecuntyEventResultinginLossOfAbility 2 HSI ConfirmedSecuntyEventmaVitalArea. 1 HAl Confirmed Security Event in a Plant Protected 1,1213141DliHUl ConfirmedSecuntyEventThatlndicatesa 234 to Reach AND Maintain Cold Shutdown EAL Threshold Value: Area , Potential Degradation in Level of Plant Safety EAL Threshold Value: 1. Intrusion into plant Vital Area by a hostile force EL.AThreshold Value: EAL Threshold Value:
I1. Loss of physical control of the Control Room due to a security OR L. Intrusion into the Protected Area(s) by a hostile force 1. A credible threat to the station reported by the NRC.
event S 2. Confirmed bomb, sabotage or sabotage device discovered in OR OR OR a Vital Area 2. Confirmed bomb, sabotage or sabotage device discovered in :2. BOTH of the following criteria are met for a credible threat
- 2. Loss of physical control of the remote shutdown capability due: the Protected Area(s) reported by any other outside agency or determined per the
- to a secunty event Safeguards Contingency Plan:
- Is specifically directed towards the station.
- Isimminent (<2 hours) tO OR
- 3. Attempted intrusion and attack of the Protected Area(s)
- 4. Attempted sabotage discovered within the Protected Area(s)
- 5. Hostage/Extortion situation that threatens normal plant
- operations HS2 Control Room Evacuation Initiated AND Plant Control CANNOT be re-established [ E]HA2 Threshold Control Room Evacuation Initiated Value:
E in 5 15 rin. 1. Entry into 2000-ABN.3200 30 "Control Room Evacuation" i EAL Threshold Value:
None 1. Control Room evacuation initiated None !
- AND U Control ofthe plant CANNOT be established in <15 mm. :
per 2000-ABN-3200.30 "Control Room Evacuation" P
Page 12 of 122ReionO Revision 0e
Ovsetr C'reek Nuetlpacr *tatan A nnpy m -vser
-- ree Anne. Exelon Nuclear TABLE PBAPS 3-1: Emergency Action Level (EAL) Matrix (Cont'd)
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS (cont.)
HA3 Natural OR Destructive Phenomena HU3 Natural OR Destructive Phenomena Affecting a Vital Area Affecting the Protected Area EAL Threshold Value: EAL Threshold Value:
- 1. Confirmed earthquake requiring reactor scram in accordance 1. Felt earthquake with 2000-ABN-3200.38 Station Seismic Event OR
- ,OR :2. Report by plant personnel of a tornado strike within the
- 2. Tornado or wind speeds > 100 mph causing damage to Plant Protected Area Vital Structures (Table H-I) OR
. OR :3. Sustained wind speeds >*75 mph as indicated by on-site
- 3. Report of visible structural damage to ANY Plant Vital meteorological instrumentation Structure (Table H-I) due to natural or destructive phenomena OR None None OR '4. Vehicle crash within the Protected Area Boundary that may
- 4. Vehicle crash damaging or affecting Plant Vital Structure potentially damage plant structures containing functions and (Table H-1) systems required for safe shutdown of the plant.
'5. Abnormal Intake Structure level, as indicated by EITHER: :5. Report of turbine failure resulting in casing penetration or Z *>6.0 ft. MSL (> 4.92 psig on PI-SWS-1[2]) damage to turbine or generator seals.
- < -4.0 ft. MSL (<0.50 psig on PI-533-1172 or PI-533- : 6. Abnormal Intake Structure level, as indicated by EITHER:
1173) >4.5 ft. MSL (>426 psig on PI-SWS-1[2])
OR MSL = Mean Sea Level < -3.0 ft. MSL
(<0.94 psig on PI-533-1172 or PI-533-1173)
HA4 Fire OR Explosion Affecting Operability of iU131, D
- U4 Fire Within themProtected Area Boundary [ T Safety Systems Required for Safe Shutdown NOTIExtinguishedin<5 15mn. of Detection F EAL Threshold Value: EAL Threshold Value:
"I. TFireor explosion causingadamage to asPlanthVital Structure ' I.Fire within or contiguous to a Plant Vital Structure (Table H-1),
(Table H-I) or affecting one or more Safe Shutdown AND
- Systems (Table H-2) Fire is NOT extinguished in _15 min. of EITHER:
AND
- Control Room notification Safe Shutdown System operability is required
- Verification of alarm OR
- 2. Report by plant personnel of an unanticipated explosion within the Protected Area Boundary
, resulting Structuresin visible damage to a Plant Vital (Table H-1)
Table H-1: Plant Vital Structures Table H-2: Safe Shutdown Systems Reactor Bldg. Main #1 EDG Vault
- Isolation Condenser Turbine Bldg.
- 4160 Safeguard Busses
- ESW Intake Structure
- Condensate Transfer 0 Service Water Page 13 of 122 Revision Oc
Ovster Creek Nuclear Sqtation Annex ý-. -I -- .- - - - --
OvYter Cul iwh-nr 5atu AnExelon Nuclear TABLE OCNS 3-1: Emergency Action Level (EAL) Matrix (Cont'd)
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS (cont.)
HAS Release of Toxic or Flammable Gases Within IH 23 4 HU5S Release of Toxic or Flammable Gases '
a Facility Structure Which Jeopardizes . Deemed Detrimental to Safe Operation Operation of Systems Required to Maintain of the Plant EAL Safe Operation OR to Establish or Maintain SEAL E Threshold TeoVl Value:
-Cold Shutdown 1. Report or detection of toxic or flammable gases that could EELThreshold Value- enter within the site boundary in amounts that can affect :
E : None None 1. Report or detection of toxic gases within Plant Vital normal operation of the plant Structures (Table H-l) in concentrations that will be life OR to plant personnel 2. Report by Local, County or State officials for potential Sthreatening
"" : OR
- evacuation of site personnel based on an offsite event C) 0 2. Report or detection of flammable gases within Plant Vital Structures (Table H- 1) in concentrations affecting the safe 1SG6 Other::H 60ter~ndtios~2Ft3hch4he Conditions Existing Which inthe : HS6 Other Conditions Existing Which in the 121l~3141 operation of the plant HA6 Other Conditions Existing Which inthe 111
__r_____
DI:nuI6 Other Conditions Existing Which inthe Judgment of the Emergency Director Judgment of the Emergency Director ,4D Judgment of the Emergency Director 2u dgterndtionseExistingyhirchth 34 Warrant Declaration of General , Warrant Declaration of Site Area Warrant Declaration of an Alert dWarrant Declaration ofan Unusual Event Emergency . Emergency EAL Threshold Value EAL Threshold Value; EAL Threshold Value: ,1 Other conditions, exist which in the judgment of the EAL Threshold Value:
01 ._ i continment containmenttbero : Emergency Director indicate actual or likely major failures ' o eiw h hjd nfh .
ment oergec Ecmtain Direor einte acua juor ie m l ofaue ajo s be degraded and that increased monitoring of plant functions Emergency Director indicate a potential degradation in the OR is warranted level of safety of the plant
- 2. Potential uncontrolled radio nuclide release, which can
- of plant functions needed for protection of the public i wre reasonably be expected to exceed I Rem TEDE or 5 Rem CDE Child Thyroid plume exposure levels at the Site Boundary Table H-I: Plant Vital Structures Table H-2: Safe Shutdown Systems Reactor Bldg Main #1 EDG Vault 0 EDGs 0 Isolation Condenser 0 ADS Turbine Bldg Transformer/Condensate #2 EDG Vault
- 4160 Safeguard Busses
- Control Room Ventilation
- CRD *0 ESW Intake Structure 0 Core Spray 0 RBCCW 0 SGTS e Containment Spray Condensate Transfer a Service Water Page 14 of 122 Revision Oc