ML040690952
| ML040690952 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 02/26/2004 |
| From: | Gallagher M AmerGen Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| -RFPFR, 2130-04-20051, NUMARC/NESP-007 | |
| Download: ML040690952 (138) | |
Text
Amer Gert AmerGen Energy Company, LLC www.exeloncorp.com An Exelon Company 200 Exelon Way Kennett Square, PA 19348 10CFR50 Appendix E.IV.B February 26, 2004 2130-04-20051 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Oyster Creek Generating Station Facility License No. DPR-16 Docket No. 50-219
Subject:
Supplement #2 to Oyster Creek Generating Station Implementation of Emergency Action Levels Developed from NUMARC/NESP-007 Methodology Refetences: (1) Letter from M. P. Gallagher (AmerGen Energy Company, LLC) to USNRC, dated March 10, 2003 (2) Letter from M. P. Gallagher (AmerGen Energy Company, LLC) to USNRC, dated December 12, 2003 (3) Conference call between AmerGen and NRC staff on January 20, 2004 This letter is being sent to supplement the proposed revision to the Oyster Creek Emergency Action Levels (EAL) and the EAL Technical Bases. In the Reference (1) letter, AmerGen Energy Company, LLC (AmerGen) submitted a change to the EAL & EAL Technical Bases based on the methodology outlined in NUMARC/NESP-007, "Methodology for Development of Emergency Action Levels, Rev. 2. In Reference (2), AmerGen proposed a change to the EAL Submittal in response to NRC comments.
AmerGen hereby submits an additional response to questions resulting from a teleconference call between NRC and AmerGen officials (Reference 3). Attachment 1 provides the NRC questions and AmerGen's response to each question. Attachment 2 contains the revised EAL Comparison Summary of Differences. Attachment 3 provides the revised Emergency Action Levels & EAL Technical Bases pages.
Ac5
OCGS Implementation of Emergency Action Levels 2130-04-20051 February 26, 2004 Page 2 The attached changes have been reviewed and agreed upon with the State of New Jersey, Bureau of Nuclear Engineering. These EALs will not be implemented until completion of the NRC's review and approval, and initial training is completed.
Lastly, AmerGen hereby requests review and approval of the revised EALs by March 10, 2004.
If you have any questions or require additional information, please contact Doug Walker at (610) 765-5726.
Sincerely, Michael P. Gallagher AmerGen Energy Company, LLC Director, Licensing and Regulatory Affairs Attachments:
- 1. Response to Request for Additional Information
- 2. Revised EAL Comparison Summary of Differences
H. J. Miller, Administrator, Region I, USNRC R. J. Summers, USNRC, Senior Resident Inspector, Oyster Creek P. S. Tam, Senior Project Manager, USNRC K. Tosch, Director, NJBNE/NJDEP File No. 03005
AT7ACHMENT 1 OYSTER CREEK GENERATING STATION Docket No. 50-219 License Nos. DPR-16 Supplement #2 to Oyster Creek Emergency Action Levels & EAL Technical Bases Response to NRC Questions
Supplement #2 to the Oyster Creek Emergency Action Levels & EAL Technical Bases - RAI Page 1 of 5 Comments to Oyster Creek EAL Conversion Review Per Teleconference with NRC on Jan 20, 2004
- 1. Summary of Differences has errors:
Barrier EAL comparisons missing (FU1, FA1, FS1, FG1)
Response: EAL comparisons for FU1, FA1, FS1, and FG1 have been added to the summary of differences document.
"RC 2" reference not correct, appears to be RCS #4 Response: RC2 reference has been revised. Additional corrections to the summary of differences document have been made.
OC EAL 1.a.1 reference to NESP 007/r2 missing (appears to be 'FC 2" or Fuel Clad
- 2)
Response: Comment incorporated.
- 2. Comment: Throughout the document, reference is made to mode numbers (1,2,3,4) which do not appear to match the OC tech. Specs. Per OC licensing, there are no corresponding mode numbers for plant conditions. Correct EALs to plant specific designations.
Response: Mode identifiers were included in the original EAL submittal, page 4, providing a cross reference of mode numbers used in the EALs to the OC mode descriptions identified within Tech Specs. In addition, the cross reference of mode numbers to OC Tech Spec mode descriptions is included at the bottom of Table OCNS 3-1.
Comment: There are numerous NESP-007, Rev. 2 references to a start-up (BWR) and hot shutdown which do not appear to be used (e.g., Modes 1, 2). Explain your modes of operation relative to your references in the EALs.
Response: Mode identifiers were included in the original EAL submittal, page 4, providing a cross reference of mode numbers used in the EALs to the OC mode descriptions identified within Tech Specs. Oyster Creek Mode 1 includes both Power operations (NUMARC Mode 1) and Startup (NUMARC Mode 2).
Comment: References to either NESP-007, Rev. 2, or NEI 99-01, Rev. 4, should be completely stated.
Response: The EAL submittal has been revised to identify those specific OC EALs where NEI 99-01, Rev 4 is used as the referenced EAL. The basis statements in the following EALs were revised to reflect specific implementation of NEI 99-01, Rev. 4:
MU6, MA6, MS5, PC 3.a.1.
Supplement #2 to the Oyster Creek Emergency Action Levels & EAL Technical Bases - RAI Page 2 of 5
RA6 declares Alert for this.
Response: RU5 IC has been clarified to state: "Potential Damage OR Potential uncovering of Spent Fuel".
- 4. Comment: RU7 - How does "10 x normal" relate to 2x TS? Add justification.
Response: Oyster Creek has taken a conservative position relative to the EAL for the ISFSI facility, in coordination with the NJ Bureau of Nuclear Engineering. Existing Technical Specification (TS) for ISFSI has a dose rate limit of 100 mrem/hr on the centerline of the Horizontal Storage Module (HSM) door. The most recent survey has a highest contact dose rate on the centerline of HSM door of 3 mrem/hr. Ten times the baseline would be 30 mrem/hr, which is below the existing TS value.
- 5. Comment: Table F-1, page 50. Typo, 30" versus 30'. Suggest spelling out the symbols
<, >, ", ' throughout the document.
Response: Typo was corrected to indicate 30 inches. For consistency and ease of use, symbols are used throughout the EALs. This is consistent with other Exelon EALs.
- 6. Comment: Table 3-2, page 51 - (1.b.1) There is no shine only indicator of fuel damage.
(Have leak + clad, and leak only, but NO clad only rad reading.)
Response: In the basis statements for 1.b.1, it is noted that another indicator of clad damage would be necessary in the event that no LOCA occurred, such as the reactor coolant activity (EAL 1.d.1). This is consistent with other Exelon sites that have adopted NUMARC/NESP 007 Rev. 2 EALs.
- 7. Comment: Basis does not support EAL for > 50 gpm leak. (Does for =1<50)
Response: The 50 gpm threshold in EAL 2.d.3 was selected based upon makeup capability through the CRD system, and capacity of the Drywell sump pumps. Leakage below 50 gpm is within the capacity of these systems, using normal operating procedures. Leakage in excess of 50 gpm will be identified through indications in the Control Room, such as Drywell pressure. The RCS barrier EAL 2.c.1, Drywell pressure and indications of a leak provide backup EAL categorization for large RCS leaks.
- 8. Comment: 3.d.2 (page 60) Question on why <3.0 psi emergency d/w blowdowns are acceptable not answered satisfactorily. Need justification or remove. Not iaw 007/2.
Response: The EAL for the Primary Containment Barrier loss, will be revised to remove reference to >3.0 psig in the EAL. The bases has been revised to state:
"Intentional venting of primary containment per the EOPs to the secondary containment and/or 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 within normal operating limits (1.1 -
1.3 psig)."
Supplement #2 to the Oyster Creek Emergency Action Levels & EAL Technical Bases - RAI Page 3 of 5
- 9. Comment: Page 62, 63 typo "should should" Response: This typo appears to be a result of a printing issue and has been corrected.
- 10. Comment: Page 67 appears to be paragraph missing at top of page.
Response: No paragraph is missing. The Basis is correct as written.
- 11. Comment: MG4 - Is > sign correct? Was cannot be determined supposed to be in this section. (Removed from -30" references, seems like it should be removed here). Page 76 seems to contradict -30".
Response: The greater than sign ">" is correct. If level is higher than -30", than it is considered greater than -30". The information on page 76, with reference to 0" TAF, is provided as a reference only and does not contradict the EAL threshold value.
- 12. Comment: MS4 - Inclusion of ARI does not still appear to be justified. Timing does not appear to be appropriate with 007/2 basis". Reference incorrect.
Response: NUMAC/NESP 007 Rev 2, basis for SS2, uses this EAL to ensure timely recognition and emergency response of a failure to scram. Failure of Alternate Rod Injection (ARI) to shutdown the reactor would be immediately recognized by the Reactor Operator.
ARI is designed to function and shutdown the reactor in less than 45 seconds. HCU scram valves open in less than 38 seconds and rod insertion is less than 5 seconds.
ARI is included in IC SA2 per NEI 99-01 as a manual scram method. Additionally, NEI 99-01 Rev. 4, SA2 indicates that failure of manual scram would escalate the event to a Site Area Emergency.
The reference to SS2 was verified to be correct.
- 13. Comment: MA4 - Mode questions and inclusion of ARI, as well as correct reference Response: Mode applicability for MA4 would only be power operation (Mode 1)
(Reference question #2).
ARI is included in the manual scram function, which is consistent with NEI 99-01, rev. 4 (Reference question 12).
Inclusion of manual scram function, EAL 2, is an additional conservatism undertaken by Oyster Creek and is a departure from NUMARC/NESP-007. This was included to provide consistency with Hope Creek, at the request of the NJ BNE.
The reference to SA2 was verified to be correct.
- 14. Comment: MS5 - Mode questions and reference Response: Refer to question 2 for Mode question response. Specific reference to NEI 99-01 has been included in EAL MS5.
Supplement #2 to the Oyster Creek Emergency Action Levels & EAL Technical Bases - RAI Page 4 of 5
- 15. Comment: MU6 - correct reference Response: Specific reference to NEI 99-01 has been included in EAL MU6.
- 16. Comment: HG1 - What is considered "remote shutdown". Use panels and systems, RAI appeared to change the logic in the EAL.
Response: Remote Shutdown is the capability to control and shutdown the reactor plant from outside the control room. Loss of Remote shutdown capability occurs when the control function of the Remote and Alternate (Local) Shutdown panels is lost. Loss of remote shutdown capability is a cumulative condition, requiring various combinations of loss of control of shutdown panels, and is associated with capabilities described with the Oyster Creek Safeguards Contingency plans. Specific combinations that would create a condition warranting declaration of a General Emergency under this EAL would be considered Safeguards material. Individuals assigned decision-making accountability, in consultation with Site security, would determine that the threshold, item 2, has been exceeded, and that a GE declaration would be made. The remote shutdown system has what is termed the remote shutdown panel, in the 480v room, and local shutdown panels in several areas of the plant, which only control certain functions. It is a combination of the remote and local shutdown panels that must be lost, along with defined target sets, which would determine or establish the loss of shutdown capability for this EAL.
- 17. Comment: HS1 - #2 doesn't match 007/2 and appears limiting.
Response: HS1 #2 refers to a confirmed bomb or sabotage within a Vital area. HS1 #1 refers to intrusion into a plant vital area by a hostile force. In both cases, the threat is to a plant vital area. NUMARC/NESP-007, rev 2, HS1 is titled "Security Event in a Plant Vital Area", and HS1 #2 under 007/2 references to site specific Safeguards Contingency Plan. The currently approved OC Safeguards Contingency Plan does not direct specific EAL classifications, but requires the Shift Manager to refer to the appropriate EP procedure for event classification. The Oyster Creek HS1 EAL IC threat level is consistent with HS1 - 007/2, in that both classify based upon threat to a Vital Area.
- 18. Comment: HU1 - #4, is not iaw 007/2. This event is an alert in HA1.
Response: HU1 #4 is "Attempted sabotage discovered within the Protected area".
Confirmed sabotage within the Protected Area would result in an ALERT declaration in accordance with HA1. Attempted sabotage is indicated by evidence of a potential degradation in the level of safety of the plant and actual degradation cannot be confirmed.
- 19. Comment: HS2 -? Does your referenced procedure prescribe transfer of functions within 15 min.? If yes, then you should specifically state this.
Response: The referenced procedure does not specify transfer of functions within 15 minutes.
Supplement #2 to the Oyster Creek Emergency Action Levels & EAL Technical Bases - RAI Page 5 of 5
- 20. Comment: HA3 - #2, 007/2 does not classify based on "damage" occurring.
Response: NEI 99-01,Rev. 4, HA3 classifies this event based upon winds exceeding a site specific value and resulting in visible damage to sample vital area structures. The existence of visible damage is intended to discriminate against lesser events. Oyster Creek will revise the NUMARC reference within HA3 to indicate implementation of the guidance of NEI 99-01, Rev. 4 for EAL HA3.
- 21. Comment: HA4 - Page 122, typo "safety" should be safe?
Response: The typo in HA4 has been corrected to indicate "Safe" shutdown.
- 22. Comment: HU4 - Damage to vital areas is not an NOUE. Not iaw 007/2.
Response: HU4, item 2 will be revised as follows:
"Report by plant personnel of an unanticipated explosion within the Protected Area Boundary resulting in visible damage to permanent structures or equipment."
Reference to Table H-1 (the list of Plant Vital Structures), will be removed from HU4, item 2. Damage to Vital Areas is classified within HA4, and is consistent with HA1 from NUMARC/NESP 007. Table H-1 remains in the basis of HU-4 because it is used in conjunction with HU4, item 1.
- 23. Comment: HA6 - TSC/OSC activation should not be an alert? Command and control of plant operation should not be leaving the control room.
Response: The basis statement for HA6 - Emergency Director Judgment will be revised to delete the following:
"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."
These deletions are to eliminate any potential confusion regarding the "staffing" of emergency facilities being used as a basis for an emergency declaration.
ATTACHMENT 2 OYSTER CREEK GENERATING STATION Docket No. 50-219 License Nos. DPR-16 Supplement #2 to the Oyster Creek Emergency Action Levels & EAL Technical Bases Oyster Creek EAL NUMARC Comparison Summary of Differences - Revised
Summary of Diffcrences NUAMARC/NESP-007 Rev. 2 to Proposed OCNS Emergency Action Levels NUMARC/
NESp-007 OCNS EAL OCNS EAL D)iffcrences/Justification E~xamp~le IC/EAL AUI.l RUI None AUI.2 RUI None r AOCNS does not have telemetered perimeters monitors, therefore this EAL is not required AU 1.4 A/A OCNS does not use automatic initiation of real time dose assessment therefore this EAL is not required AU2.I RU6 None AU2.2 R U5 None AU2.3 RU7 None A U2.4 RU2 None AA1.l RA I None AA 1.2 RA I None
.3 AT1 OCNS does not have telemetered perimeters monitors AA 1therefore this EAL is not required AA 1.4 A/M OCNS does not use automatic initiation of real time dose assessment therefore this EAL is not required AA2.I RA5 None AA2.2 R,15 None AA2.3 RA6 None AA2.4 RA5 None AA3.I RA2 None AA3.2 RA2 None ASIA RSI None AS 1.2 A/A OCNS does not have telemetered perimeters monitors therefore this EAL is not required AS 1.3 RSI None AS 1.4 RSI None AG1.1 RGI None AG!1.2 AT/A OCNS does not have telemetered perimeters monitors therefore this EAL is not required AG 1.3 RGI None AG I.4 RGI None l[l U I lIU3 None 11 U 1.2 IIU3 None HIU 1.3 11U3 None
-lUI.4 11U3 None Unanticipated explosions covered Linder IIUJ (NUMARC I-IU 1.5 11U4 I-U2) since explosions are more logically associated with OCNS threshold l-U4 I IUl.6 IIU3 None Summary of Difflerences OCNS-NUMARC Diff Revc 1.doc
Summary of Differcnces NUNMARC/NESP-007 Rev. 2 to Proposed OCNS Emergency Action Levels NUNIARC/
NESP-007 OCjVSEAL OCNS EAL l)iffcrences/.Justification Example IC/EAL U 1.7 HU3 Added site specific high and low intake water levels as other conditions appropriate for this IC I1U2 11U4 Included unanticipated explosions from HIU 1.5 IlU3.1 IIU5 None lHU3.2 IU5 None This EAL threshold has been written to conform with IC IH1U4 IHIU41 I
ul Zregarding devices as amended and endorsed by the NRC in a U.1 Iletter from Mr. B. A. Boger to Ms. lynette Hlendricks (NEI) dated 2/4/02.
I-IU4.2 HIUI This wording conforms to the criteria of [HU4 as amended and I approved by NRC for post-9/l I security issue resolutions IUl U51U6 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"
[IA 1. I 1IA3 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 analvsis.
This EAL is written in accordance with NEI 99-01, rev. 4, in classifying this event based upon exceeding site specific wind HIA 1.2 11/13 values and resulting visible damage to sample vital area structures. The existence of visible damaige is intended to discriminate against lesser events.
I-IA 1.3 I1113 None HI A 1.4 11,13 None I-IA 1.5 1143 None HA 1.6 A/4 No plant safety equipment is potentially impacted by missiles generated by turbine rotating failures at OCNS.
I-IA 1.7 1I3 Added site specific high and low intake water levels as other conditions appropriate for this IC lHA2 1114 None H1A3.I IIS None I1A3.2 IA5 None IHA4.I 11,11M None H1A4.2 HlA I None HIA5 IIA2 None I IA6 11A6 None I-IS I. I 111 None IHS 1.2 1181 None Summary of Differences OCNS-NUMIARC [)iff' RevOe 1.doc
Summary of Diffcrenccs NUMNlARC/NESP-007 Rev. 2 to Proposed OCNS Emergency Action Levels NUNIARC/
NExSP-00e7 OCNS EAL I)iffcrences/.Justification E-xamlel~l IC/EAL IS2 HS2 None IIS3 11S6 None IGI1 lIG1 None I-G11.2 HG!
None HG2 1HG6 None SUl fU!
None SU2 AIU9 None The condition "In the opinion of the Shift Supervisor, the loss of the annunciators or indicators requires increased S U AIUf) surveillance to safely operate the unit(s)" has not been 5U MU6 included in the condition consistent wvith changes to IC SU3 in NEl 99-0 1, Rev. 4. This statement does not provide useful assessment criteria to the EAL threshold.
The MODE applicability [I,2] is a deviation from NUMARC SU4.I RU3
[all] in that, the SJAE Radiation Monitor, selected as an 'other
,U4.
Rindication' will only be a valid indication of Fuel Clad Degradation mode's [ 1. 21.
SU4.2 RU4 None "Pressure boundary Ieakage' not applicable to OCNS since no SU5 AIU7 distinction is made between unidentified or pressure boundary leakage in the OCNS Technical Specifications.
SU6 HU8 None SU7 A1U3 None SAlI MAl 2 None Added "Loss of manual SCRAM capability indicated by SA2 iAt f 4 failure of ALL manual SCRAM methods to achieve reactor shutdown" per resolution of NJ 13NE concerns and consistency with I-lope Creek Station.
SA3 ilASf None The condition "In the opinion of the Shift Supervisor, the loss of the annunciators or indicators requires increased 5A4 A If surveillance to safely operate the unit(s)' has not been included in the condition consistent with changes to IC SU3 in NEI 99-01, Rev. 4. This statement does not provide useful assessment criteria to the EAL threshold.
SA5 A IA I None SSI ALSI None SS2 AIS-None SS3 XMS3 None SS4
,1/.$5 Implements NEI 99-01, Rev. 4 13WR specific criteria.
Revision 2 of NUMARC/NESP-007 simply specified loss of Summary of Differences OCNS-NUMARC Difl'RevOel.doc
Sunmmary of D)ifferenccs NUMARC/NESP-007 Rev'. 2 to Proposed OCNS Emergency Action Levels NUMARC/
NESP-007 OCiVSEAL OCNS EAL D)ifferences/hJustification Example IC/EAL
[site-specific function] necessary to maintain Hot Shutdown.
Revision 4 of NEI is specific in defining this condition for BWRs as inability to maintain parameters below lleat Capacity Temperature Limit.
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 %vas not included in this SS5
,1S7 EAL.
Added the condition "OR CANNOT be determined' consistent with OCNS EOPs for loss of ability to determine water level.
RPV water level must bc assumcd to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
SS6 Al6 None Added the condition "OR CANNOT be determined" consistent wvith OCNS EOPs for loss of ability to determine water level.
SG]
GII RPV wvater level must be assumed to be below the barrier threshold if RPV wvater level cannot be determined by any direct or indirect method.
Added the condition "OR CANNOT be determined' consistent wvith OCNS EOPs for loss of abilit! to determine wvater level.
SG2.1 11iG4 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.
SG2.2 AiG-4 None I'UI I
UI None FAI IAI None FS I I-SI None IG I FG1I None I'C. I
.(t.I None Added the condition "OR CANNOT be determined" consistent I with OCNS EOPs For loss ofabilitv to determine water level.
F C.2 1.a.2 RPV wvater level must be assumed to be belowv tile barrier threshold if RPV water level cannot be determined by any direct or indirect method.
"AND Indication of RCS leak inside drywellF criteria added to IC.3 Ih. I clarify intent and distinguish from loss of containment cooling events which can manifiest itseif symptomatically similar to RCS leakage wvhich is the intent of the IC.
Summarv of Differences OCNS-NUMARC Diff RevOc I.doc
Summar) of D)iffercnces NUNIARC/NESP-007 Rev. 2 to Proposed OCNS Emergency Action Levels NUMARC/
NExSmP-007l OCNS EAL OI)fferences/.Justification E~xample IC/EAL re.4 A'M No 'other' fuel clad loss/potential loss indicators identified for FC.4 N/A OCNS FC.5 I. fI None RC. I 2.. 1/2/3/4 None RC.2 2.c. I None RC.3 2.h. 1 None Added the condition "OR CANNOT be determined' consistent with OCNS EOPs for loss of abilitv to determine water level.
RC.4 2.. I RPV water level must be assumed to be below the barrier threshold if RPV wvater level cannot be determined by any direct or indirect method.
RC.5 MAl No 'other' RCS loss/potential loss indicators identified for RC.5 N/A OCNS RC.6 2.f I None PC. I
- 3. c. 12/3 None 3.e. I PC.2 3..it 11/23 None PC.3
- 3. h. 1 None The Revsion 2 NUMARC example ElA prescribes an RPV water level in conjunction wvith the Maximum Core Uncovery I'CA4
- 3. a I Time Limit (MCUTL). This is a misapplication of the PC.4 3.a.1 MCUTL, which was corrected in revision 4 of NEI 99-01.
Primary Containment Flooding required (Entry into SAMG) is now specified in the current NUMARC document.
PC.
AVA No 'other' PC loss/potential loss indicators identified for PC.5 V/A OCNS PC.6 3.f I None Summary of Dilflrenees OCNS-NUMARC DifI' Rev0e I.doc
ATTACHMENT 3 OYSTER CREEK GENERATING STATION Docket Nos. 50-219 License No. DPR-16 Supplement #2 to the Oyster Creek Emergency Action Levels & EAL Technical Bases Emergency Action Levels & EAL Technical Bases (For information only)
Ovster (Creek Niielptir Sta-tinrl Arnnex Fvtelnn Nivvlptr flvt'r ('rp'k Niwlonr Sttinn Aiunov F.rolnn N1,,clpqr Oyster Creek Nuclear Station Annex Section 3 Emergency Action Levels (EALs)
EAL Technical Bases Revision Of Prepared Iy:
Operations Support S icres, Inc.
1716 White Pond L.ane Waxhaw, NC 28173 (704) 243-0501 www.ossi-net.com Prepared lor:
EIelon Nuclear 200 Exclon Way Kennett Square. PA 1934N Purchase Order 0 () 042079
Ovster Creek Nuclear Station Annex Exelon Nuclear Section 3: Classification of Emnergencies 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 meetin rg 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 wvere 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 (I) 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 Emeruencv Action Levels are determined, declaration will be made on the high0est classification level for the Unit.
3.1 Enicrgeancv 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 Ermer-encv Action Levels. Mode Applicability provides the unit conditions when the Emergency Action Levels represent a threat. 'T'he Basis provides definitions of terms, explanations and justification for including the Initiating Condition and Emergency Action Level. I)efinitions 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
\\'alues. These values are based on the criteria established Uinder Revision 2 to NUNMIARC/N ESP-007, "Methodology for D)evelopment 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 ProdUlct Barrier Degradation
- System Malf'unctions
- I lazards andl Other Conditions Page 2 of 123 RRevision Of
Chcfor CrtPU Nvirlonr 15%tntinn Avinpy Irxvcln Nneucl.ar
-1 I------ _. --
EAL Tlhresilold Values are sorted tinder common Initiating Conditions (ICs). These ICs can be Symptom-or Event-based, and applicable to all or only designated Operational Conditions / Modes OPCONs. Th1e Initiating Conditions (IC) and associated EAL Threshold Values are summarized in the [AL 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-Categorics, and appropriate Mode designator provided:
- A bnoriial Radliologgical Levels /EfflueLtis ("R')
Radiological Effluents Abnormal Rad Levels Coolant Activity Irradiated Fuel Accidents Fissioni Product Barriers ("F')
Fission Product Barrier ffatrix comiipriseLd of:
FuIel Clad Barrier Reactor Coolant System (RCS) Barrier Primarv Containment (PC) Barrier S;stemn Malfulntionls ("Al')
Loss of AC Power Loss of DC Failure of RPS Decay I leat Removal Loss of Annunciators RCS Leakage / RPV Draindown Loss of Communications Technical Specifications Hlazarls a(nd Otlher Conditions ("11")
Security Events Control Room Evacuation Natural or Man-Made Events Fire / Explosion Toxic or Flammable Gas I)iscretionary Page 3 of 123 RRevision of
Ovstcr 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 I through 4 are based on Reactor Mode Switch Position and average reactor coolant temperatUre. "Defiueled" Mode was established for classification purposes Linder NUMARC/NESP-007 to reflect conditions where all fuel has been removed from the Reactor lPressure Vessel.
MODE)
TITLE COND)ITION I
lPower Operation Technical Specification de fiition 2
I-lot Shutdown Shutdown conI(litiofl or Refuel Mode as defined by Technical Specifications and Reactor Coolant Temperature not below 212 degrees F or not vented.
3 Cold Shutdown Technical Specification Definition 4
Refueling Technical Specification definition of Refuel Mo(le and Reactor coolant temperature below 212 dearees F and vented.
I)
I)eflueled No fuel in the Reactor Vessel The EAIL Matrix is designed to provide an evaluation of the Initiatine Conditions from the wvorst 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 wvill be inder classified. All Recomgition Catcgories should be reviewed flor applicability prior to classification.
An appropriate EAI, numbering system is provided as a user aid. ICs are coded with a two letter and one number code. For example: IIA
'T'he first letter is the Recognition Category designator. In this case, 11 stands for "I lazards and Other Conditions". The second letter is the Classification Level: LU" for Unusual Event, "A" for Alert, "S" for Site Area Elmergency, and (G" for General Emergency. Thle nulnber is a sequential number for that Recognition Catevorv series.
All Initiatingy Conditions, which are describing the severity ofla common condition (series). wvill have tile same number (e.g. I IA I. I IA2. etc.).
Page 4 of 123 RRevisionl Or
Ovstcr Creek Nuclear Station Annex Exelon Nuclear A Fission Product Barrier (FPB) Table is provided as a subset to the Recognition Category 1" (FPB Degradation) of the EALN 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 POTEN'TIAL LOSS).
3.2 EAL Technical Basis Table OCNS 3-2 serves as the Technical Basis for the EAL Matrix.
Tile table consists of the following sections for each Initiating Condition (IC), sorted by Recognution Category:
Initiating Condition Threshold Valuc(s)
Mode Applicability Basis Plant-specific References D)ilferences (deviations from NUMARC/NIE'SP-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/N ESP-007.
3.3 General EAL Implemientation ]Philosophy A broad spectrum of discretion in classifying events is provided in the "I)iscretionary" categ ory' under lHazards 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. EIRO members should be mindflul than an approach is needed wvhich is conservative wvith 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 fior 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. AlthoLughl the consequences ordeclaring 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 flor the public health and safety are appropriate. B3ecause of Page 5 of 123 RRevision Of'
(_)vc.tir CrpiL Nm-1v-irqtnttnn Annov FvdeAnn Mig-lenr OvI----
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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.
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 tile 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 belowv the ElALs 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 emergencv 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 emergencv classi fications.
Page 6 of' 123 e
Rievisionl Of
U--lrn XT-1-o A+ASA PAA IU9lAIAA BOAS A_ A __A Oyster Creekc Nuclear Station Annex TABLE OCNS 3-1: Emereeney Action Level (EAL) Matrix GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ABNORMAL RAD LEVELS / EFFLUENTS RG1 Actual or Projected Site Botundars 2 t RSI Actual or Projected Site Boundary Dose RAI Release > 200 X ODCM Limit for RUI Release> 2 X ODCM Limit for Using Actual Meteorology:
Using Actual Meteorology:
> 15 mini l
Jil2Jl3lJ D
60 min.
> IOO mRem TEDE
> I0OmRem TEDE EAL Threshold Value EAL Threshold Value:
OR OR I. Unplanned radiological release lasting> 15 min. in excess of I Unplanned radiological release lasting> 60 min. in excess of
> 5000 niRem CDE Child Thyroid
> 500 mRem CDE Child Thyroid Table RI "Alert" threshlolds Table Rl "Unusual Event" threshold EAL Threshold Valte EAL Tlhreshold Value:
AND AND I Radiological release in excess of Table RI "General I.Radiological release in excess of Table Ri "Site Area Releases CANNOT be detennined in S 15 mi. (from time Releases CANNOT he determined in S60 mi. (from tim
= Eergency" threshold Emergency" threstold Table RI threshold was exceeded) to be below Table R2 Table RI threshold was exceeded) to be below Table R2 S
AND AND "Alert" thresholds "Unusual Event" thresholds Releases CANNOT be determined in < 15 min. (from time Releases CANNOT be determined in < 15 min. (from time OR OR f Table RI threshold was exceeded) to be below Table R2 Table R I threshold was exceeded) to be below Table R2 "Site 2 Unplanned radiological releases lasting > 15 ma. in excess of 2 Unplanned radiological releases lasting > 60 min in excess of Dose Assessment "General Emergency" thresholds Area Emergency" thresholds ANY Table R2 "Alert" threalold 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 la-Plant Radiation Levels Impede Plant RU2 Rise In Plant Radiation Levels by a Operations
[IlZl3lJlDl Factor of 1000 by 2lEliL Di I EAL Threshold Value:
EAL Threshold Value:
I Radiatone readings a 15 mR/hr in EITHER of the fallnsmg'
' I Valid area radiation monitor readings indicate an unplanned a
Main Control Roos rise by a factor of 1000 over normal levels as detected by OR either permanent or temporarily installed radiation monitors or
-f NoeNone Central Alarm Station b
aulsre N:oOR by manual suwey i
E 22 In-plant radiation readings > I R/htr in areas requiring access RU3 High Off-gas Radiation Levels n
in order to maintain safe operation or perform a safe EAL Threshold Value:
E t I 2
shutdown I
Off-gas radiation reading > Hi-Hi alarm value for > 15 min.
RU4 Higls coolant activityE IZ E a;
None
'5 None
- EAL Threshold Value:
-N.
Reactor coolant activity > 0.2 pCi/gm DE]
Table Rl -- f~uent Monitr ThresholdsTable R2 Dose Assessment Thresholds Tabe R At or besond the Site Boundarr based on a I hou rclease duration Release PointlMonitor General Emergency Site Area Emergency Alert Unusual Event Method General Emergency Site Area Emergency Alert Unusual Event Main Stack RAGEMS
>200XODCM 4.6.11.4
> 2 X ODCM 4.. 61 A4 MiStcRAESSample N/A N/A OR OR Torus/Dry1e1t 2" Vent
> 21 Ii iCi/cc HRM
> 2.1 pCi/cc HRM
> Ct'S A
> CPS U Analysis
> 200 X ODCM 4.6.1.1.5
>2XODCMR4.6..1.5 via SBGT Rteactor tldg. via SOGT
> 3. 11pCi/cc FIRM
> 0.3 pCi/cc IIRM
> CPS A
> CPS U Field Team O
> I 0RemhrWhole Body
>100 mRembrWhole Body
> IO mRemAppticalle oniorig
> 501(1 nRem CDE Child Thyroid
>500) mRem CDE Child Thyroid
> 34 miem CDE Child Thyroid Turbine Bldg. RAGEMS Via EFI-4 & EFI-33
> 3.0 pCi/cc HRM
> 0.3 lCi/cc HRM
> 200 X Hi-Hi
> 2 X Hi-Hi alarm setpoint alarm setpoint Dose
> t )(( siRenm TEDE
> t i() utRem TEDE
>1). I )mRem/hMTEDE Asseasmentn OR OR OR OR
> 500O soRem CDE Child Thyroid
> 5011 mRem CDE Child Thyroid
> 34 mRem COE Child Tbyroid
>0(.34 mRem CDE Child Thrroid Service Water Effluent N/A N/A
> 3.13 E4 cpn
> 4.15 E2 cpm Plant Modes; Powe Operations Hot Shntdowns (h 212'F)
'Fl Cold Shutdown (l 212 "Fl Refuel Defueled HRM= High Range Monitor LRM= Low Range Monitor Page 7 of 123 Revision Of
"--fo 1--ol'
%Nlseloar Aufa Ann-E-elan Nuclear TABLE OCNS 3-1: Emereenev Action Level (EAL) Matrix (con't)
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT ABNORMAL RAD LEVELS / EFFLUENTS RA5 Major Damage OR Uncovering of BU11 I
' RUS Potential Damage OR Potential uncovering 2 3 4 D Spent Fuel of Spent Fuel EAL Threshold Value:
EAL Threshold Value:
NNone Valid unanticipated Hi alarm on one or more Refuel Floor
- 1. Uncontrolled water level drop in the Spenit Fuel Pool with None ARMs (Table R-3)
ALL irradiated faiel 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 WaterLevel That H.a OR Will ((El]
l RU6 Uncontrolled Water Level Drop in a
Un. cover Irradiated Fuel ut dte Reactor Reactor Cavity Cavity EAL Threshold Value:
=.
None'nEAL Threshold Value:
. Unexpected Skimmer Surge Tank Lo-Lo level alarm None None
- 1. Report of visual observatios 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 Isalto EAL Threshold Value:
I Radiation readings> 10 times normal at ANY of the following ISFSI locations:
None None None on contact with roof OR on contact with shield door centerline OR on contact with shield wall Table R-3: Refuel Floor ARMs C-5, Crit Mon C-9, North Wall C-IO, North Wall l
B-9, Open Floor Page 8 of 123 Revision Of
n-vf-r V-ee Nee Qts.*i-Annex huelon Nuclear t-[C -
I.,t~a -
llll 11
.asn TABLE OCNS 3-1: Emereency Action Level (EAL) Matrix (Cont'd)
FISSION PRODUCT BARRIER MATRIX (Applicability: Modes 1 & 2 ONLY)
L12L IJ FISSION PRODUCT BARRIER STATUS FGI: GENERAL EMERGENCY FSI: SITE AREA EMERGENCY FAI: ALERT FUI: UNUSUAL EVENT Fuel Clad - LOSS X
X X
X X
X Fucl Clad - POTENTIAL LOSS X
x x
X Reactor Coolant SvStes - LOSS X
X X
X X
X Reactor Coolant System-POTENTIAL LOSS X
X X
X Primar,, Containment - LOSS X
X X
X X
X Primam Containmelt - POTENTIAL LOSS X
X
- 1. FUEL CLAD BARRIER
- 2. REACTOR COOLANT SYSTEM BARRIER
- 3. PRIMARY CONTAINMENT BARRIER LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS I, RPV le, el < 0" TAF (not intentionally a RPV lttcl I I V <3 A
lowered by procedure) tn Entry into SAMGs as required by a.RPV Water Level C.
RPV level <-30" TAANO OR None None E
s CANNOT be determined R
I Containment Hi-Range Radiatiot I.
Cootainment Hi-Range Radiation I Containment Hi-Range Radiation
- b. DryweRi Monitoring System (C4RRRMS)
None Monitoring System (CHRRMS)
None None Monittrng System (CHRRMS)
Monitorinag
> 440 R/hr
>45 R/thr I02.OE+4 R/hr I Rapid, unexplained drop in dryoell I. Otyn ell pressure >3.0 psig pressure following an initial rise
- c. Drymell Pressare None None AND None
- 2. ORy elsep~eoc.site
- 3. Dsywiellpressure 044 psig Indication of RCS leak inside Drywell Nvith LOCA conditions indicating a containment breach IICool >300 (DEI) olde MainSteamLinebr 3I RCS Leakage > S0 gpm I Failure of all isolation val es in ANY I. Coolant activity 0 300 pCi/gat (DEI)
I.Uiout aoSemLn ra
.ESLaueS psone tine penetrating Primnsoe outside containroent OR Containment to close whenrarequired OR
- 4. Unisolable prinmary sy stem leakage AND
- 2.
Unisolable tsolaion Condenser tube outside of drvsvell as indicated by Dosvnstrearn pathbas exists to rupture exceeding EITHER of the folloving in environment rpueore ormore reas requiringauscrm:
OR MG-32()(1 1t Max Normal
- 2. Intentional renting per EMG-32()().0(2 is remperature required.
- d. Breached / Bypassed None OR OR None EMG-32()(111
) Max Normal
- 3. Unisolable primary system leakage Radiation Level outside of drvssell as indicated by exceeding EITHER of the folloiing in one or more areas requiring a scram:
- EMG-3200.1It Max Nurmal Temperature OR
- EMG-3200.1It Max Normsat Radiation Level
- e. Coistaisesent I. Containment Ha, 6
u.Ž Hydrogen None None None None None AND Concenltation Containment Or-iu 2Ž5%
- f. Emergesscy Direclor I. ANY condition in the judgment of the Emergency Director that indicates Loss or I. ANY condition in tltijtidgment of tIe Emergency Director that indicates Loss or I. ANY condition in the judginent of the Emergency Director that indicatesLoss Judgasent Potential Loss of the Fuel Clad barrier.
Potential Loss of the RCS barrier or Potential Loss of the Primary Containment barrier Page 9 of 123 Revision Of
I__
--sla Nw TAVBtLE -II CN
- cler Acttion LevelAnn Mix
,C t-TABLE OCNS 3-1: Ernerpency Action Level (EAL) Matrix (Cont'd)
GENERAL EMERGENCY l
SITE AREA EMERGENCY I
ALERT I
UNUSUAL EVENT SYSTEM MALFUNCTIONS MGI Prolonged Loss of ALL Offsite AC l l ll MSI Loss ofALL Offsite AC Power AND t t l
MUl o of Al ffe Poe to Essential Power AND Prolonged Loss of ALL Loss ofALLOnsiteAC Power to E E2M A
reduced to a spblge powtersource bosgesa Ire relL Bosses for Greater Than 15 Min.
Ii Onsite AC Power Essential Busses hnELTrsodVle EAL Threshld Value:EAL Threshld Value:minutes sutch that any additional single failure would resitlt ELTrsodVle inL stationVa blackouthol Vaue
. Loss of offsite power to BOTH 4160V Busses IC and I BOTH4160V Busses 1C and ID de-energizedfor> 15 min I. BOTH 4160V Busses IC and ID de-energized for> 15 min ID for> 15 min.
AND EAL Threshold Value:
ANY oftie following:
I Lossofoffsitepowerto BOTH 4160V Busses IC and ID for Restoration of at least one emergency bus within I hour 15 min.
is not likely AND RPV level CANNOT be maintained > 0" IAF OR EITHER of the 4160V busses IC or ID de-energized for a
CANNOT be determined
> 15 min,.
X Torus water temperature and RPV pressure exceeds the MAI Loss of All Offsite Power AND Loss of All l.1 I1 Heat Capacity Temperature Limit (Figure F. EMG-Onsite AC Power to Essential Busses Dur Cold Shutdown Or Refueling Mode EAL Threshold Value:
I BOTH 4160V Busses I C and I D de-energized for > 15 min.
MS3 Loss of All Vital DC Power l
Ml Unplanned Loss of Required DC EAL Threshold Value:
[ TM~j Power During Cold Shutdown or LIEJL None I. Loss of ALLvital DC power indicated by < 115VDC None Refueling Mode > 15 Min.
indication on 125 VDC Busses B and C for > 15 min.
EAL Threshold Value:
,. LossofALLvital DCpowerindicatedby<115VDC indication on 125 VDC Busses B and C for > 15 min.
MG4 Auto and Manual SCRAM NOT LI li MS4 Auto and Manual SCRAM NOT
[
MA4 Auto SCRAM NOT Successful Successful, AND Loss of Core Cooling or Successful EAL Tlireshold Value:
Heat Sink EAL Threshold Valuc:
EITHER EAL Threshold Value:
I RluS setpoint for all automatic SCRAM exceeded E
I RPS setpoint for an automatic SCRAM exceeded AND I
Rl'S setpoint for all automatic SCRAM exceeded
~
AND Failure ofautomatic RPS, ARI and manual SCRAM to D
Failure of autonatic RPS, ARI and manual SCRAM to reduce reactor power < 2%
Failure of automatic SCRAM to achieve reactor shutdown reduce reactor power<2%
OR 2
None D ITHER
,2.
Loss of manual SCRAM capability indicated by failure of ALL ANDEIHR
'l On manual SCRAM attempts to achieve reactor shutdown a
RPV level CANNOT be restored and maintained i
w ic
> -30" TAF OR CANNOT be determined a.,
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 FI]2]
EALThreshold Value:
[l3l o
EAL Threshold Value:
M1. Unplamned loss of all Tecinical Specification required systems E
I Tos water temperature aed RPV pressure CANNOT he available to provide decay heat removal functions
'a u
maintained below the Heat Capacity Temperatlre Limit AND None (Figure F, EMG-3200.02)
Uncontrolled temperature rise that approaches or exceeds 212 IF None Page 10 of 123 Revision Of
(-f 0-ste Qfee all f;-Rstn A.- ~
V-I-a N. 4 TABLE OCNS 3-1: Emergency Action Level (EAL) Matrix (Cont'd)
GENERAL EMERGENCY SITE AREA EMERGENCY
-ALERT UNUSUAL EVENT SYST EM MALFUNCTIONS (cont.)
MS6 Inability to Monitor a Significant Transient MA6 Unplanned Loss of Most or All Safety System 1 2 MU6 Unplanned Loss of Most or All Safety System In Progress EF2 Anttociation or Indication in Control Room Annunciation or Indication in the Control IiiiIl l EAL Threshold Value With Either (l) aSignificant Transient in Progress, or(2)
Room > 15 Min.
E I.
A significant transient is itt progress (Table M- )
Compensatory Non-Alarming Indicators are Unavailable EAL Threshold Value:
AND EAL Threshold Value:
1.Unplannedloss,for>1l5min.,ofMOST(Note l)orallof
~
ALL of the following are lost:
I. Unplanned loss, for > 15 min. of MOST (Note I) or all of EI.sER:
Safety system annunciators (Table M-2)
EITHER:
Safety system annunciators (Table M-2)>
None Safety function indicators (Table M-3)
Safety system annunciators (Table M-2)
OR Plant Process Computer OR Safety function indicators (Table M-3)
Safety functiots indicators (Table M.3)
AND EITHER:
A significant plant transient is in progress (Table M-I)
OR Plant Process Computer is tuaavailable MS7 LossofWaterLevel itileReactorVessel I
MU7 RCS Leakage That Has or Will Uncover Fuel in the LL3 l
4 1
PALFThresholdMValte:
None Reactor Vessel Notte M
Unidentified leakage > 10 gpm t
I EAL Threshold Valse:
OR I
RPV level < " TAF OR CANNOT be determined 2
Identified leakage > 25 gpmn
~MU8 Unplanned Loss of ALL Onsite OR 24 Offsite Communications Capabilities o
EAL Threshold Value: l a
None None None I.
Loss of all onsite communications (Table M-4) affecting the 0 =
ability to perform routine operations OR U
l'2.
Loss of all offiite communications (Table M-4)
MU9 Plant is not brought to required operating mode within Technical Specifications LCO tE 2M Action Statement Time None None None EAL Threshold Value:
- 1. Required operating mode is NOT reached within Tech. Spec
,s CO actton completion time Table M-l: Significant Plant Transients Scram
> 25% thermal power change Sustaited power oscillations (30 watts/cm' LPRM peak to peak)
Stuck Open EMRVs ECCS Injection TakieM-2: Safety System Annunciators table M-3: Safely Fs,,cstiou Indicators ECCS (B, C)
Reactor Poaer, Pressure and Les el Containment Isolation (G, H. J)
(Panel 4F, 5F, PF)
Reactor Scram (G)
. DecaNy Fleat Renoal (Panrel I F/2F)
Process Radiation Monitoing (IltF)
Cooatneot Safety Fancann (Panel I IF, 12XR, 16R)
Table M4: Communications NOTE I Onsie Offsite Offsite MOST' refers to a loss of -75% or a significant
. Plant Paging Conventional tel. lines ED Hotline risk that a degraded plant condition could go System
- Cell Phones
. New Jersey State Police (NJSP undetected.
Use is not intended to require a
. Conventional Radio Notification Lise) detailed count of annunciators/indicators.
telephone lInes ERF Ocean County Notification Line
. Cell Phones
. ENS
. NJ State ED Hotline
. Radio
. HPN
. Ensironsental Assessment Direct Bureau of Nuclear Engineering Line Information Line Page 11 of 123 Revision Of
Ov-t-C-eek N-l1-Ssn Annex ExYelon Niielefir v-tc I-
-OM -o "9 slo
,tl--l sI~i TABLE OCNS 3-1: Emergencu Action Level (EAL) Matrix (Cont'd)
GENERAL EMERGENCY SITE AREA EMERGENCY l
ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS HGI Security Event Resulting in Loss Of Ability HSIFF1 liSt Confirmed Security Event inaVital Area.
F l
HAI Coifinned Security Event in a Plant Protected iHUI Confirmed Security Event That Indicatesa to Reach AND Maintain Cold Shutdown II.I+/-II EAL Threshold Value:
LLLLLI Area L..LLLLJ Potential Degradation in Level of Plant Safety lll.
J EAL Threshold Valne:
I Intrusion into plant Vital Area by a hostile force EAL Threshold Value:
- EAL Threshold Value:
I. Loss of plysical control of the Control Room due to a security OR I. Intrusion into the Protected Area(s) by a hostile force I. A credible threat to the station reported by the NRC.
event
- 2.
Confirmed bomb, sabotage or sabotage device discovered in OR OR OR a Vital Area
- 2. Confirmed bomb, sabotage or sabotage device discovered in j2. BOTH of the followitg 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 security event 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 />)
- 83. Attempted intrusion and attack of the Protected Area(s)
,4. Attempted sabotage discovered within the Protected Area(s)
OR i 5. Hostage/Extortion situation that threatens normal plant operations HS2 Control Room Evacuation Initiated AND i
HA2 Control Room Evacuation Initiated Plant Control CANNOT be re-established LLI1.1.1 ItrEAL Threshold Value:
in *15 min.
I. Entry into 2000-ABN-3200.30 "Control Room Evacuation" EAL Threshold Value:
^
None
- 1. Control Room evacuation initiated None 0
AND lO Q
Control of tie plant CANNOT be established in < 15 min.
'n per 2000-ABN-3200.30 "Control Room Evacuation" Page 12 of 123 Revision Of
tdv1--~
0-op, V-1-~su Q--fi Annex Fxenln Nuclear le 11.1-
---~ta --
lll..
.~s....
TABLE PBAPS 3-1: Emergency Action Level (EAL) Matrix (Cont'd)
GENERAL EMERGENCY SITE AREA EMERGENCY l
ALERT UNUSUAL EVENT HAZARDS AND OTHER CONDITIONS (cont.)
HA3 Natural OR Destructive Phenomena
- H1U3 Natural OR Destructive Phenomena Affecting a Vital Area
[
1 l2l3lJl D
l Affecting the Protected Area l1I3J EAL Threshold Value:
EAL Threshold Value:
I. Confirmed earthquake requiring reactor scram in accordance I 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 mnph causing damage to Plant Protected Area Vital Structures (Table H-I)
OR 4
'3. Sustained wind speeds > 75 mph as indicated by on-site a3.
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 Use Protected Area Boundary that may 5
- 4. Vehicle crash damaging or affecting Plant Vital Structure potentially damage plant structures containing functions and 3
o (Table H-I) systems required for safe shutdown of the plant.
a OR OR aE
- 5. Abnormal Intake Structure level, as indicated by EITHER:
'5. Report of turbine failure resulting in casing penetration or
> 6.0 ft. MSL (> 4.92 psig on PI-SWS-1 [2])
damage to turbine or generator seals.
OR OR
. -4.0 ft. MSL (<0.50 psig on P1-533-1172 or PI-533-6 Abnormal Intake Structure level, as indicated by EITHER:
> 4.5 ft. MSL (> 4.26 psig on PI-SWS-I [2])
(< 0.94 psig on P1-533-1172 or P<-533-1173) 13 HA4 Fire OR Explosion Affecting Operability of r
HU4 Fire Within the Protected Area Boundary Safety Systems Required for Safe Shutdown
[ tl23 J
NOT Extinguished in < 15 min, of Detection hi Z2JIfl?]
EAL Threshold Value:
EAL Threshold Value:
- a.
Fire or explosion causing damage to a Plant Vital Structure I Fire within or contiguous to a Plant Vital Structure (Table H-I)
(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 inanticipated explosion within the Protected Area Boundary resulting in visible damage to permanent
'___:__structures or equipments.
T A
.bleH-i; Plant Vital Struotures Reactor Bldg.
Main SI EDGVat Turbine Bldg.
Transformec/Condnsatc
- 2 EDG ValtI Control Room Complex Transfer Pad EDG Fuel Oil Storage Tank Intake Structure Table H-2: Safe Shutdown Systems EDGs Isolation Condenser l
ADS 4160 Safeguard Busses Control Room Ventilation SDC (IC& ID)
SGTS Containment Spray Condensate Transfer Service Water Page 13 of 123 Revision Of
Ovyter Creek Nuclear Station Annex FV-l-n N-l-ee TABLE OCNS 3-1: Emergency Action Level (EAL) Matrix (Cont'd)
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT HAZARDS AND OTIER CONDITIONS (cont.)
,HA5 Release of Toxic or Flammable Gases Withi HU5 Release of Toxic or Flamnmable Gases a Facility Structure Which Jeopardizes
[1J2j31JID:
Deemed Detrimental to Safe Operation Elililil]
Operation of Systems Required to Maintain of the Plant Safe Operation OR to Establish or Maintain EAL Threshold Value:
0E Cold Shutdown I Report or detection of toxic or flammable gases that could EAL Threshold Value:
enter within the site boundary in amounts that can affect Notte None I
Report or detection of toxic gases within Plant Vital normal operation of the plant Structures (Table H-I) in concentrations that will be life OR threatening to planit personnel
- 2. Report by Local, County or State officials for potential OR evacuation of site personnel based on an offsite event
- 2 Report or detection of flattttable gases withtn Plant Vttal Structures (Table H-I) in concentrations affecting the safe HG6 Other Conditions Existing Which in the
.HS6 Other Conditions Existing Which itl the
'HA6 Other Conditions Existing Which in the MiTTT) 11116 Other Conditions Existing Which in the 24 Judgment ofthe Emergency Director Judgment of the Emergency Director t 2 3 4 0 Judgment of the Emergency Director Judgment of the l
Director Warrant Declaration of General Warrant Declaration of Site Area Warrant Declaration of an Alert Warrant Declaration of an Unusual Event Emergency Emergency EAL Threshold Valse EAL Threshold Value:
EAL Threshold Valie:
.I Other conditions, exist which in the judgment of tle EAL Threshold Value:
I I Actual or imminent core degradation with potential loss of a
t w
i Emergency Director indicate that plant safety systems may I. Other conditions exist which in the judgment of the e
containment E ergcy Diret inicate be degraded and that increased monitoring of plant functions Emergency Director indicate a potential degradation in the OR actal rwikeyrajotfilueslevel of safety of the plant 5
°4. OR
' ~~~~~~~~of plant functions needed for protection ofthle public saatdlvlosft fh ln 2 Potential sncontrolled radio nuclide release, which can o p Q
reasonably be expected to exceed I Rem TEDE or 5 Rem CDE Child Thyroid plume exposure levels at the Site Boundary I able H-1: Plaut Vital Stnicftames Reactor Bldg.
Main
- I EDO Vault Turbine Bldg.
Transformer/Condensate 12 EDG Vault Control Room Complex Transfer Pad EDG Fuel Oil Storage Tank Intake Structure Table H-2: Safe Shutdown Systems EDGs Isolatiots Condenser l
ADS 4160 Safeguard Busses Control Room Ventilation SDC (I C & I D)
CRD ESW Core Spray RBCCW SGTS Containment Spray Condensate Transfer l
Service Water Page 14 of 123 Revision Of
Ch-ctor Crpek Nivelt-irqtntinii Avinvy Evvinn Nnelp'ir
-1...... -
lal)ec 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVE'LS / EFFLUENTS Many 'ALs are based on actual or potential degradation of fission product barriers because of the increased potential for of'site 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 l'r emergency classiification.
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, ofl7site radiological conditions may, result wvhich require off'site 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.
RadioloL ical ffiluients 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 off'site field measurements indicate doses or dose rates above classifiable limits.
Abnormal Rad Level Sustained gencral 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 Activitv Elevated activity in this process stream may be indicative ol' fucl clad degradation and is considered to be a precursor oflmore serious problems.
Elevated coolant activity in excess of'Tlchnical 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 uncoverv of' spent fucl in cither the Spent lFuel Pool or Reactor Cavity. This subcategory also addresses incidents associated wvith ISISI.
Page 15 of' 123 RRevision 01'
A.-py Flr non NlT.ol.nio t
L'k l
I I
J I'
1 t1 %
II CII 1t I AI II f1 B
A 111t.*
.ttC 1 I..I 1
Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EIFLIJENTS RGI INITIATING CONDITION Actual or Projected Sitc Boundar' Dose Using Actual Meteorology:
> 1000 mRcnm TEDI Oft
> 5000 mrRem CDE Child 'I'hiroid EAL THRESHOLD VALUES
- 1. Radiological release in cxcess ofl'l)le Rl "Gencral Emer gecylC " thresiold ANI)
Release CANNOT be determined in < 15 min. (from time 'Table RI threshold was exceeded) to be below Table R2 "'Gcneral Enmergency" thresholds
- 2. Radiological releases exceed ANN' Table R2 'General Emiergencv-threshold.
l'aI)le RI: Effluent Monitor Thresholds Release Point Effluent Mionitor General Emergency Torus!D)rvwell 2" Vent Main Stack RAGEMS I IRM
> 21.0 pCi/cc via SI3GTto Main Stack Reactor 13Building via Main Stack RAGEMS I IRM
> 3.0 pCi/cc SBGT to Main Stack Turbine 13uiilding via 1E 11-4 & 1.1 1-33 to T3 TB Stack RAGEMS I IRM
> 3.0 pCi/cc Stack I IRM= I ligih Range Monitor I'age 16 of 1 23 R
llevisioll Of
Ovster Creek Nuclear Station Annex Exelon Nuclear T'ab)le 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
A1BNORMNIAL RADIATION LEV'E'lS / I;FFLIJl'INTS RGI (con't)
EAL THRESHOLD VALUES (con't)
Table R2: D)ose Assessment Tb resholds Metho(d General nI<mergency Sample Analysis N/A Field Team
> 1000 mRem/hr Whole Bodv Mlonitorino*
i to,>
> 5000 mRcm C[)l, Child Thvroid I)ose Assessment*
> 000 rRern '1.['
> 5000 mRcm Cl)l Child TIhyiroid
- At or beyond Sitce Boundarv based on a I hour release duration MODE APPLICABILITY AL.L, BASIS Site Boundary - As specified in the OI)CM.
Total E.ffective D)ose Fquivalent (TlElD.) The sum of the deep dose equivalent (for external exposuirc) and the committed effective dose equivalent (for internal exposure) and 4 days ol'deposition exposire.
Committed Dose Eiquivalent (CDE) T'he I)ose Equivalent to organs or tissues of'relfcrence that will be received from an intake of'radioactive material by an individual during the 50-year period folloving the intake. The 5000 mRem integrated child thyroid dose was established in consideration of'the 1:5 ratio of the lEPA Protective Action Guidelines lor 'T'E[DE and child Thyroid Committed l)ose E'quivalent (C[)l,). Actual meteorology is used, since it gives the most accurate dose projection.
TIable RI:
'filcilt Mtonitors - Classification is based oln the instantancoLIs release rate value if'NO dose projections can be performed or verified wvithin 15 minutes of meeting or exceeding the specified Release Rate valie.
Monitor indications (rounded) are calculated using the computerized dose model RAC wvith l.OCA source terms applicable to each monitored pathway in conjunction with annual average (low end) metcorology and a one-hour release duration For Torls!D/DW 2" ventinm via S13GT or Turbine Building, Vent and seven-hour release flor Reactor B3uilding ven tig via SBGT. The inputs arc as Follows:
lPage 17 of' 123 RReision SO1or
Ovster Creck Nuclear Station Annex Table 3-2: OCNS EAL Technical Basis Exclon Nuclear RECOGNITION CATEGORY (R)
ABNORMAL RAI)IATION LEVELS / EFFLUENTS RGI (con't)
BASIS (con't)
Stability Class Wind Speed Wind I)irection (from)
Accident Release Rate Release D)uration Time after S/[)
Torus/D)W Vent Main Stack I) 1 3.0 mph 31(6 (SE,)
I.OCA 20.9 pCi/cc I hour I hour Reactor B3uildimne Vent Main Stack 1) 13.0 mph 3160 (SE)
LOCA 2.99 pCi/cc 7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> I hoLur Trurbine BuildiLjn Vent'TB Stack I) 6.0 mph 3 160 (SE)
LOCA 3.08 pCi/cc I hour I houLr Table R2:
Field Team Monitoring - The values are for surveys or iodine air samples taken at or beyond tile SITE BOUNDARY and are thle most accurate indicator of the condition. Field data are independent of release elevation and meteorolo-v. The assumed release duration is I hour. 1)irect readin-iodine monitors are not available. Sampling of'radioiodine by adsorption on charcoal media fiollowed by field analysis are used for determining the iodine valIe.
REFERENCE(S)
I. 6630-ADM-40 10.03 Oyster Creek E-mergency Dose Calculation Manual
- 2. "Review ol'Calculations to Support l.ALs RGI and RS I" dated I 1/22/02 to 1P.
Thompson From G.
Seals. Radiological Engiineer
- 3.
'iNlaximnlm1 EAL Levels F-flluent Radiation Monitors dated 12/22/95 to T. BlIount from J. Stevens -
E'nuineer. E.ngineering & D)esi NUMARC IC AGI D)IFFERENCES
- 1. NUMARC IC AG 1.2 - OCNS does not have telemetered perimeter monitors. therefore this EAL. is not required Page 18of' 123 R
Revzisionl Of
0,Lctgr Crivk Nvit-h-ir Stnfinn Annov F-Yplnn lN'ielpesir (b Itp
- -ua
'tp. r Ktntnn rAn FlrdAn Nuwap'*r Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
A13BNORMNAL RAD)IATION LEVELS / EFFLUENTS RS]
INITIATING CONDITION Actual or Projected Site Boundary )ose Using Actual Meteorology:
> 100 mRem Tl'l)E OR
> 500 rnRem Cl)E Child T'hyroid EAL THRESHOLD VALUES
- 1. Radiological release in cxcess of'Tat)le I "Site Area Emergecyilc" thrcshold ANI)
Release CANNOT be determined in < 15 min. (from time T'able R I threshold was exceeded) to be below Table R2 "Site Arca Emergencv" thresholds OR
- 2. Radiological releases exceed ANV 1'able 1Z2 "Site Area Emergency' threshold Table RI: Effluent Monitor Thresholds Release Point Effluent Monitor Site Arca Emergency Torus/Drvwvell 2' Vent Main Stack RAGEMS llRM
> 2.1 LtCi/CC via Sl3GT to Main Stack M
Reactor Building via RcacT tor Main Siac Main Stack RAGEMS lHRM
>0. 3 pCi/cc Sl3G'l to M\\>ain Stacl;k Turbine Building via TB Stack RAGEMS lIRM
> 0. 3 pCi!cc H 1 &-
& 1, 1-33 to 1T13 Stack
-IRNM= l-lih Rance Monitor Page 19 of' 1R9o3 Revisionl of
0-.,v.ttr CrPA-Nnv1P-ir.qt-itinn Annex E~xtlon Nuclca-r Table 3-2: OCNS EAL Teelinchical Basis RECOGNITION CATEGORY (R)
A13NORZMAL RAI)IATION LEVELS / EFFLUENTS RSI (con't)
EAL THRESHOLD VALUES (con't)
Table 1R2: D)ose Assessment Thresholds Methiod Site Area E mergeicy Sample Analysis N/A Field Team
> 100 mRem/hr Whole l3odv Mfonitorin-*
ORz i rin*
> 500 mRcm CI)A Child 'I'hyroid
> I100 mRem T1,1' I)ose Assessment
> 100 ORz
> 500 mRem CDI-Child Thyroid
- At or bevond Site Boundary based on a I hour release duration M1ODE APPLICABILITY A1,1, BASIS Site Boundary - As specit'ied in the O[)CM.
Total Eff'ective l)ose lEquivalent (TEDE) The sum of the deep dose equivalent (for external cxposure) and the committed effective dose equivalent (f'or internal exposure) and 4 days of'deposition exposure.
Committed l)ose Ecquiv'alent (CDl) '[he l)ose Equivalent to organs or tissues of'reference that wvill be received from an intake of'radioactivc material by an individual during the 50-year period followving the intake. T'he 500 mRenm inte,!rated child thyroid dose w"as established in consideration of the 1:5 ratio of the E'.PA l'rotectivc Action Guidelines flor lTDE-and child Thyroid Committed Dose l'quivalent (CDE-). Actual meteorology is used. since it gives the most accurate dose projection.
Table R 11:
Effluent Monitors - Classilication is based on the instantaneous release rate vlueic if'NO close projections can be performed or verified within 15 minltes of meeting or exceeding the specific( Release Rate value.
Monitor indications (rounded) are calculated using the computerized dose model RAC with l,OCA source terms applicable to each monitored pathway in conjunction with annual average (low end) meteorology and a one-hour release duration l'or Torus!D[W 2" venting via S13GI' or Turbine Building \\Vent and seven-hour release lor Reactor Building Vlenting;, via SBGT. The inpUts are as follows:
Page 20 of' 123 R
ReIsiC
- Sor1O
Ovster Creek Nuclear Station Annex Table 3-2: OCNS EAL Tcchnical Basis Exclon Nuclear RECOGNITION CATEGORY (R)
ABNORMAL RAI)IATION LEYELS / EFFLUJENTS RSI (con't)
BASIS (con't)
Stability Class Wind Speed Wind Direction (from)
Accident Release Rate Release l)uration l'ime after S![)
Torus/DW Vent Main Stack I) 13.0 mph 31606 (SE)
LOCA 2.09 pCi/cc I hour I hour Reactor ffuildino Vent Main Stack D
13.0 mph 3 160 (S1.)
LOCA 0.299 pGCi/cc 7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> I hlour Turbine Buildinu Vent 1'13 Stack I) 6.0 mnplh 3160 (SE)
LOCA 0.308 pCi/cc I hour I hour Table R2:
Field Team Monitorinu, - The values arc for surveys or iodine air samples taken at or beyond the SITE 13OUNDARY and are the most accurate indicator of the condition. F'ield data are independent of release elevation and meteorology. The assuIned release duration is I hour. l)irect reading iodine monitors are not available. Sampling of radioiodine by adsorption on charcoal media flollowed by field analysis are used for determininm, the iodine value.
This event wvill be escalated to a General Emergency when actual or projected doses exceed El'A-400-R-92-001 Protective Action Guidelines per IC RG1.
REFERENCE(S)
- 1. 6630-A1)M-40(10.03 Oyster Creek Emergency Dose Calculation Manual
- 2.
- Review of Calculaltionls to Support EALs RGI and RSI dIlated( 11/22/02 to 1'. Thompson from G.
Seals. Rladiological Enuineer
- 3. 'iMaximum E'AI. Levels Efluient Radiation Monitors" dated 12/22/95 to T. Mlount from J. Stevens -
Enizineer. Enuineerino & I)esi-n NUMARC IC ASI DIFFERENCES
- 1. NUMARC IC ASI.2 - OCNS does not have telemetered perimeter monitors, therefore this EAL is not required.
Page 21 of 123e llevisioll 01'
Ovster Creek Nuclear Station Annex Exclon Nuclear Tal)e 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RAI)IATION LEVELS / EFFLUENTS RAI INITIATING CONDITION Release > 200 X Ol)CM limit lor > 15 min.
EAL THRESHOLD VALUES
- l. Unplanned radiological release lasting > 15 min. in excess ol Table RI "Alcrt' thresholds ANI)
Releases CANNOT be determined in < l 5 min. (from tim ITable R I threshold was exceeded) to be below Table RZ2 "Alert" thresholds OR
- 2.
Utiplanned radiological releases lasting > 15 minutes in excess ol ANY Table R2 "Alert" threshold Tahle RI1: Effluent Monitor Thresholds Release Point/Nionlitor Alert Main Stack Noble Gas RAGEIM1S Ioru /Df)r1s eII 2" Vent % ia SIl(;GT
> CI'S A Reactor BlId.. i ha Sll(
> CI'S A
- Turbine Bldg. Noble Gas RAGEMNS V*ia.FI-4 & :1 1-33
> 200 X Ili Ili alarin selpoint Service W\\'ater
> 3.13 I4 cpi Table R2: D)osc Assessment Threshioldls Mlethod Alert
> 200 X ODCM 4.6.1.1.4 Sample Analysis OR
> 200 X ODCM 4.6.1.1.5
> I0 mRem!hr Whole B3ock Field leam Monitoring*
> 34 mRcm CDl Child lIhvyroid
> 10 mRem!hr l1lEl)l I)osc Assessmlent' OR
> 34 mRcem C)lE Child 'I'h)roid
- At or beyond Site B3oundary based on a I houLr release duration Page 22 of 1 23 s
Revzisionl Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
A1BNORMNIAL RAD)IATION LEVELS / IFF'LUENTS RAI - Cont'd MODE APPLICABILITY AIL BASIS Unplanned - Not the result of' an intended Cvolu1tion and reqUiring corrective or mitigative actions Sustained - A sustained unplanned release of this greater magnitude that cannot be terminated in 15 minlutes represents an uncontrolled situation that is an actual or potcntial substantial degradation of the level ofsafety of'the plant. The degradation in plant control implied by the filct that the release cannot be terminated in l S minutes is the prinmary concern. The Emergency Director should not wait until IS minutes has elapsed, but should declare an event as soon as the release is determined to be uncontrolled or projected to be noin-isolable within 15 minutes.
T'al)le R1:
Effluent Monitors - Monitor indications are calculated based oln tile mctlhodology of'the site Offsite Dose Calculation Mlanual (ODCM). The li Il i alarm set point for the Turbine l3iilding gaseous effluent monitor is set conservatively to indicate when a potential release may approach OD)CM 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 l.Al, requircs tile magniftide of the monitor reading to be two hundred times thle set point, sustained for >1 5 minutes. and assessment bv a dose projection indicating an off'site dose rate in excess of two hundred times OI)CM limits. The Main Stack value (CI'S A) is calculated on a periodic basis consistent with the multiple of'200 times ODCM limits.
Service Water efllUCelt monitor value of 3.13 E4 cpm is consistent xvith 200 times O[)CM 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 Ol)CM f'or greater than 15 minutes.
Samples Analysis - The calculation called flor in this EAL should also be conducted whenever a liquid release occurs for which a radioactive discharge permit wvasn't prepared or that exceeds the conditions on the permit (e.g. minimum dilutioll. alarm set points, etc).
Field Team nMonitoring - The valuies are flor surveys or iodine air samples taken at or beyond the SITE B3OUNI)ARY and are the most accurate indicator of the condition. Field data are independent of release elevation an(d meteorology. The assumned release duration is I h1our. Direct reading iodine monitors are not available. Samplino of'radioiodine by adsorption onl charcoal media f6ollowed by field analysis are used for determininig the iodine value.
Page 23 of' 123 Reision; jOr0
Oster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical B3asis RECOGNITION CATEGORY (R)
A13NORMNIAL RAI)IATION LEVELS / EFFLUE'NTS RAI - Cont'd BASIS - Cont'd Dose Assessment - This lAI, includes a 15 minute average l'or the dose projection with the release point radiation monitor above two hundred times the Ifli Ili 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 'i'El)l, is calculated by dividing the yearly allowable Ol)CM limit (500 mRemlyr.) by the number of hours per year (8760 hr/yr.). and then multiplying by a factor of 200 times OI)CM T'DI,
= 200x(Tech Spec Liimit)/(hours per year)
= 200(500 mltem/yr.)!(8760 hr/yr.)
= I 1.4 mRem/hr (rounded to 10 mR/hr)
The CI)E Thyroid is calculated by dividing the yearly allowable OI)CM limit (1500 mRcm/vr.) by the number of hours per year (X760 hr/yr.), and then multiplying by a factor of 200 times OI)CM.
Cl)l.
= 200x(Tech Spec Lirnit)!(hours per year)
= 200(1500 mnleri/yr.)!(8760 hr/yr.)
= 34.2 m lem!hr (roundled to 34 mR/hr)
Releases in excess of I I mnler/lhr TED)E, 34 mRem CDE Tlhroid. or 200 times the OI)CM limits that contillnIe fIor > 15 minutes represent an uncontrolled situation and hence a potential degradation in the level of saf'ety. The primary concern is the final integrated dose 1100 times greater than the Unu1sUal Event] and the degradation in plant control implied by the flact that the release was not isolated within 15 minutes.
This event will be escalated to a Site Area Emergency wvhen actual or projected doses are determined to exceed IOC R20 annual average population exposure limits per IC IRSI.
REFERENCE(S)
- 1. 2000-ADM-4532.04 Oyster Creek Ofnfsite Dose Calculationl ManlLii
- 2. "Maximum EAI.A Levels EfIfluent Racliation Monitors" dated 12/22/95 to T. 131Bount from J. Stevens -
IEn-ineer. 1En-ineerin- &
c)esimn NUMARC IC AAI DIFFERENCES I. NUMAItC IC AA 1.3 - OCNS does not have telemetered perimeter monitors, therefore this ["Al, is not required.
- 2. NOMARC IC AA 1.4 - OCNS does not use automatic initiation of real time close assessment therce'ore this E'AI, is not reqluired.
Page 24 of' 123e Itevision1 0f'
CWctor Croviz, Nvit-lenr qtntinn Annt-v Fel'nn Nuicrlear TlaI)e 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
A1BNORMNIAL RAI)IATION LEVELS / EIFFLUENTS RUI INITIATING CONDITION Release > 2 X ODCM limit for > 60 mrin.
EAL THRESHOLD VALUE
- 1. Unplanned radiological release lasting > 60 min. in excess ol'Talble RI "Untisuial Evenit" threshold ANI)
Releases CANNOT be determined in < 60 min. (from time Table RI threshold was exceeded) to be below Table R2 -UJnustial Event" thresholds OR
- 2. Unplanined radiological releases lasting> 60 mm.in excess of'AN' Tabl)le R2 Uiinstial E-ventt" threshold T'able R1I: Effluent Monitor Thresholds Relcase Point/Mon itor I imstial Event Main Stack Noble Gas RAG ENIS Torrtzil)r iwell 2" Vent i i:i SIMT(
> CPS 1I Reactor vldg.
ia Sll(;Il
> CIS 11
- Turbinc Bldg. Noble Gas RAGENIS Via ;l't-4 & 11F1-33
> 2 X I I i I I i alairm setrxpint Service WX'ater
> 4.15 1:2 cpni Tlal)le R2: I)ose Assessment Thresholds Method Unusual Event
> 2 X ODCM 4.6.1.1.4 Samplc Analysis ORz
> 2 X ODCM 4.6.1.1.5 Field Team Mlonitoring1*
N/A
> 0.10 mR/hr 1T1)'I I)ose Assessment*
> 0.34 mRem Cl)l Child 'I'lThroid
- At or bevond Site Boundary based on a I hour release duration Page 25 of 123 Rei{CSiOI1 01'
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORNIAL, RAI)IATION LEVELS / EFFLUENTS RUI - Cont'd MODE APPLICABILITY AU, BASIS Table RI:
IEflulent Monitors - Monitor indications are calculated based on the methodologTy of the site Offsite D)ose Calculation Manual (O)CNI). The I Ii I-li alarm set point for the Turbine l3iilding -aseous effluent monitor is set conservativeli to indicate when a potential release may approach OI)CM 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 EAI requires tile magnitude of the monitor reading to be two times the set point. sustained for >60 minutes, and assessment by a close projection indicatinig an ollfsitc dose rate in excess of two times ODCM limits. TIle Main Stack value (CPS U) is calculated on a periodic basis consistent with the multiple ol'2 times Ol)CM limits. Service Water effluent monitor valuc ofl. 15 [ 2 cprn is consistent with 2 times Ol)CM allowed concentrations.
Table 1R2:
It is intended that the event be declared as soon as it is determined that the release will exceed two times Ol)CM for izrcater than 60 minutes.
Sample Analysis - The calculation called for in this EAI should also be conducted wvhenever a liquid release occurs for which a radioactive discharge permit wasn't prepared or that exceeds tile conditions on tile permit (e.g. minimum dilution, alarm set points, etc).
Dose Assessment - This [l.A, includes a 60 minute averag(e lor the dose projection with the release point radiation monitor above two times the Ili Il i alarm set point value for the entire 60 minutes. It is not intended that the release be averaged over 60 minutes, but exceed threshold lor 60 minutes.
The TleDl is calculated by dividing the yearly allowable Ol)CNI limit (500 mRcm/yr.) by the number of hours per year (8760 hr/yr.). and then multiplying by a factor of 2 times the Ol)CM limit.
T1 I I)
= 2x (eCCh1 Spec lIlmit)/(llours per ycar)
= 2 (500 mRen/vr.)!(8760 hr/yr.)
= 0. 11 mR1le!hr (rounde1Cd to 0. 10 mR/hr)
The Cl)I. is calculated bv dividing the yearly allowable O[)CM limit (1500 mRile/vr.) by tile number of hours per year (8760 hr/yr.). and then multiplying by a factor oi2 times Ol)CM limit.
CI)l
= 2x (Tech Spec Limit)!(hours per year)
= 2 (1500 mRer/yr.)!(8760 hr/yr.)
= 0.342 mRemr/lr (rounided to 0.3 4 mR/hr)
Page 26 olf 123 e
Revision 01'
Ovqter Creek Ntielear Station Annex Ejxplon Ni'lenc.r Ov~erCr~'kNnccn Sttin Ane F~lo Niiclenr~.7 Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
A13NOlRMAL RAI)IATION LEVELS / EFFLUENTS RUI - Cont'd BASIS - Cont'd Releases in excess ofO. I0 10mRern/lr 'T'1)1, 0.34 mRcm!hr C[)l' Child Thyroid, or 2 x Ol)CM 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.
REFERENCE(S)
- 1. 2000-AI)M-4532.04 Oyster Creek Olfsite Dose Calculation Manual
- 2.
'Maximurn I'Al, iLevels ffiluent Radiation Monitors' dated 12/22/95 to T. B3lount from J. Stevens -
Engineer. Engilinecring & I)esign NUMARC IC AUI DIFFERENCES
- 1. NUMARC IC AU.3 - OCNS does not have telemetered perimeter moniitors. tierefore this EAI is not required.
- 2. NUMARC IC AU 1.4 - OCNS (foes not use automatic initiation of real time close assessment therefore this l-IAL is not required.
Page 27 of' 123 Rev ision 1 O'
Ovster Creek Nuclicear Station Annex Exelon Nuclear Table3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORNIAL RADIATION LEVELS / EFFLUENTS RA2 INITIATING CONDITION In-I'lant Radiation Levels Impede Plant Operations EAL THRESHOLD VALUES
- 1.
Radiation readings > 15 mlhlr in EITHER oflhe l'ollow\\ing:
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 shutdowvn MODE APPLICABILITY ALL BASIS This EAI, addresses elevated radiation levels that impede necessary access to operator stations, or other areas containing eqluipment that must be operated manually in order to maintain safe operation or to perform a safe shutdowvn. The concern ofthe 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 ofthe level of saf'et: of the plant. The cause of the rise in radiation levels is not the major concern of this [AI,. For example. a dose rate of 15 mRAhr in the control room or high radiation monitor readings may also be indicative ofThiglh dose rates in the containment dule to a COCA. In this latter case, the fission product barrier table may indicate a SAE or GE.
Threshold \\'alue I - The X'alue of l5 iiReni/hr is derived from the general design criteria (GI)C) value of 5 RE-M in 30 cday's with ad justIlient for expected occupancy times. Although Section 111.D).3 ofNUREG-0737 "Clarification ofT'Ml Action Plan Requirements" provides that the 15 mRern/lhr 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 AlERRT.
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' 123R Revxisionl 01'
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RAI)IATION LEVELS / EFFLUENTS RA2 - Cont'd BASIS - Cont'd Threshold \\'alue 2 - Areas requiring infrequenit access and dose rate values are based on those specified in the IEmergency Operating Procedure (EMG-3200.1 1) Secondary Containment Control.
hlhe single value of RI/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 CMR 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. D)ose rates > I R/hlr may impede necessary access.
REFERENCE(S)
- 1. 17MG-3200. 1 I Secondary Containment Control NUMARC IC AA3 DIFFERENCES None Page 29 ol 123e lacevisionl Of'
Oster Creek N'uclcar Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
AB3NORMIAL RAI)IATION LEVE`lLS / EFFLUENTS RU2 INITIATING CONDITION Rise in Plant Radiation Lcvels By a Factor of 1000 EAL 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 A1,1, BASIS Unplanned - Not the result of an intended cvolution and requiring corrective or mitigative actions.
Normal L.evels - Normal radiation levels can be considered as the higlhest 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:
Aln instrument channel check indicating that the monitor has not flailed:
Indications on1 related or redundant instrumentation, or Direct measurement by plant personnel.
Classification of an UNlJSU)AI IVI.NT is warranted as a precursor to more serious events. The concern of this ['Al, is the loss of control of radioactive material representing a potential degradation of the level of safctv of the plant. 'T'he Threshold Valuc tends to have a long lead-time relative to a radiological release and thus the threat to public hcalth and safety is very low.
Area radiation measurements should include both permanent and temporarily installed area radiation monitorin, instruments as well as manual area radiation surveys but excludes Containment Hligh Range Radiation Monitors (CH-IRRMS). Increases in area radiation in normally high radiatioii areas would be classified uinder Al. RA2.
REFERENCE(S)
None NUMARC IC AU2 I)IFFERENCES None Pagc 30 of 123 Revzision1 Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATECGORY (R)
ABNORMAL RAI)IATION LEVE'LS / IEFFLUENTS RU3 INITIATING CONDITION Ili(gl Off-gas Radiation Levels EAL THRESHOLD VALUES Off-eas radiation > Ili-Ili alarm value for> 15 min.
MIODE APPLICABILITY:
- 1. 2 BASIS:
The steam jet air ejector discharge (oftfgas) radiation monitor RN051'(F') in the Control Room wvould be one of the first indicators ol' a potentially degrading core.
lie high-hligh alarm is nominally set below the Technical Specification limit. '[hc [Hi lHi alarm results in the closure of V-7-31. V-7-29 and OG-AOV-001A (00113) to isolate the off-pas 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. 'T'his indicator of elevated activity is considered to be a precursor of more scrious problems. The Technical Specification limit reflects a degradin, or degraded core condition.
If operator action to reduce off-gas radiation levels is not successliil within the 15 minuite time frame. this level of steam line activity is assumed to be indicative of the release of' gap activity to the coolant. Thle mechanics that caused oflf-as radiation to rise to this level indicate there are a degradation of lFuel Clad integrity and thus a threat to the Fluel Clad Barrier.
'I'his l.ALI is NO'T' intende(d to apply to cases caused by resin intrusion or other known factors.
REFERENCE(S)
- 1. 2000-AB3N-3200.26. Increase in Off Gas Activity
- 1. The MODI)' applicability 11.21 is a deviation from NUMARC tall] in that, the SJAI.F Radiation Monitor will only be a v'alid indication of Fuel Clad D)egradation in MOI)E's [I, 21.
Page 31 of 123 Rei{Ctsion1'O
Ovster Creek Nuclear Station Annex Table 3-2: OCNS EAL Tech nical Basis Exelon Nuclear RECOGNITION CATEGORY (R)
ABNORNMAL RAI)IATION LEVEILS / IEFFLUIE NTS RU4 INITIATING CONDITION I li-h Coolant Activity EAL THRESHOLD VALUES Reactor Coolant activity > 0.2 uCi/gm l)EI MODE APPLICABILITY:
A LL BASIS: (References)
Coolant activity in excess of Technical Specifications (> 0.2uCi/gmn) is consideredl to bc a precursor of more serious problems. The lTechniical Specification limit reflects a de-rading} or de(raded core condition. This level is chosen to be above any possible short duration spikes tinder normal conditions. An Unusuial lEvent is only wvarranted when actual IfLel clad damage is the cause of the elevated coolant samrple (as determined by laboratorv conFirmation). I lowever. fulel clad damage should be assumned to be the cause of elevated Reactor Coolant activity unless another cause is knowvn. c.v.. Rcactor Coolant System chemical decontamination evolution (during shutdown) is ongoing with resulting high activity levels.
This event wvill be escalated to an Alert when Reactor Coolant activity exceeds 300 ptCi/gm Dose Equivalent Iodine 131 per lission Product B3arrier Matrix FC. I.
REFERENCE(S)
- 1. OCNS Technical Specifications Section 3.6.A NUMARC IC SU4 DIFFERENCES None Pace 32 of 123 llevision Or
"-+--
NN-1--
Q#-+;.-
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Fr'n'o1n lN'ut#nwr
- ,&u, t1 LL. h I
lr I
A IIUI-I 1,11t II r-nI **t.-
I
.l.l.t I lable 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RAD)IATION LEVELS / EFFLUENTS RA5 INITIATING CONDITION Major D)amage OR Uncovering of Spent Flucl EAL THRESHOLD VALUES I.
\\Valid unanticipated lI-i alarm on one or more Refuel Floor ARMs (T'able R-3)
ORl
- 2.
Report of visual observation that irradiated fuel in the Spent Fuel l'ool or F'uel Transfer Canal is uncovered Table R-3 Reftiel Floor ARN~s C-5.
Crit Mon
- C-10. North Wall C-9.
North Wall 13-9.
Open Floor AMODE APPLICABILITY All.
BASIS: (References)
Off'site doses during these accidents wvould be well below the lI'A l'rotcctive 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 I-li alarm set point. which wv'ere expected during a planned evolution. should not cause an Alert to be declared.
This IC applies to spent lIel requiring water coverage and is not intended to address spent fuel wvhich is licensed for dry storag+/-e (ISIFSI). which is addressed in RU7.
NURI-G-081 8, "I'mer-ency Action levels for Light Water Reactors." forms the basis for this E AL,. The areas wherc irradiated fuel is located forms the basis for Table R-3.
Unoexpected radiation levels.
signilicantly higher than the normal background wvill generally indicate a fuel handling accident or loss of water coverin-the irradiated fuel. Readings may be from refuiel floor Area Radiation Monitors or taken during a qualified radiological survey.
Pace 33 of 123O llcevisionl or
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGOIV (RZ)
ABNORMAL RAI)IATION LEVELS / EFFLUENTS RA5 - Cont'd BASIS: (References) - Cont'd There is time available to take corrective actions, and there is little potential for substantial fluel damage. In addition, NUR['(G/CR-49S2. "Severe Accident in Spent Iluel Pools in Support of Generic Safetv Issue 82." Julv 1987. indicates that even if corrective actions are not taken, no prompt fatalities are predicted, and that riskl of injufry is low.
REFERENCE(S)
- 1. 2000-RAP-3024.01 1 1' NSSS Alarm Response Procedures NUMARC IC AA2 DIFFERENCES None Page 34 of 123 R
Rei{CrsioI 01'
Ovster Creek Nuclear Station Annex Table 3-2: OCNS EAL Technical Basis Exelon Nuclear RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RU5
.. I-.....
I C O.N....
IO.N...
I INITIATING CONDITION-Potential Damage OR Potential 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 wvill 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 123 Revision Of
OVster Creek Nuclear Station Annex Exclon 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-08 18, "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)
Ovster Creek Nuclear Station Anncx Exclon Nuclear Tablc 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS /EFFLUENTS RU6 Uncontrolled Water Level Drop in Reactor Cavity 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 [wvhen in refueling and flooded uip wvith the gates removed]
w~ill result in a control room annunciator Skimmer Surge Tank Level Lo-Lo Alarm. This alarm is validated wvith visual observation of a lowering Spent Fuel Pool level. If thle 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 ev'ent 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 wvarranted as a precursor to a more serious event.
In light of Reactor Cavity Seal failure incidents at twvo different PWRs and loss of water in'tle Spent Fuel Pit/Fuecl Transfer Canal at a IBWR 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 wvarranted 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 123ReionO Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / EFFLUENTS RU6 - Cont'd REFERENCE(S)
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (R)
ABNORMAL RADIATION LEVELS / 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
,MODE 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 tinder Category R, "Abnormal Radiological Levels / Effluents."
Page 39 of 123 Revision Of
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
Ovster 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.
- Primarv 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 123 Revision Of
(11vEast e Cree&k Nuclear Station Annex Rypinn Nuirclar Ovt'r Crook Niw1nr 5 tntiAn Annc
EYPIAH Niu'Irir 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 I 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 powver operations, it will be an active Technical Specification Action Statement. However, during accident conditions, this will represent a breach of Primary Containment.
Page 42 of 123 Revision Of
Ovster Crcek Nuclear Station Annex Exelon Nuclear 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. Drywell Radiation Monitorine
- 1. Containment [-li-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 ItCi/gm (DEI)
NA
- e. Containment HyIvdrogen Concentration NA NA
- f. Emereency 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 123 Revision Of
OvstcrCreekl Nnlelar 9tqtinn Annoy Excplon Nuclcar
, a.......
_..... S, n
-E Nula_
Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER DEGRADATION
- 2. RCS BARRIER
- a. RPV WVater Level
Oil CANNOT be determined POTENTIAL, ]LOSS NA
- b. Drvwvell Radiation Monitoring
- 1. Containment Ili-Range Radiation Monitoring System (CIIRRMS) >45 RIhr
- c. Drvwvell Pressure
- 1. Drywell pressurc >3.0 psig ANI)
Indication of a RCS Icak inside drywecl
- d. Breached / Bvrnassed
- 1. Unisolablc Main Steam Line break outside containment OR
- 2.
Unisolable Isolation Condenser tube rupture NA NA
- 3. RCS leakage >50 gpm Oil
- 4. Unisolable primary system leakage outside of drywvell as indicated by exceeding
[triIER of the following in one or more areas requiring a scram:
EMG-3200.1 I Max Normal Temperature OR EMG-3200.1 I Max Normal Radiation Level
- e. Containment I lvdropen Concentration NA NA
- f. Emergency Director Judgment I. ANY condition in the judgment of the Emergency Director that indicates Loss or Potential Loss of the RCS barrier Page 44 of 123 Revision Of
nvetor GrrppL Nivelonr iSttion Annexp F.Yelon Nuclt-ar Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER DEGRADATION
- 3. PRIMARY CONTAINMIENT BARRIER L.OSS POTENTIAL,LOSS
- a. RPV Water Level NA
- h. Drvwvell Radiation Monitoring NA
- 1. Containment lli-Range Radiation Monitoring System (CIIRRMS) > 2.0E+4 R/hr
- c. )rvivcll Pressure
- 1. Rapid, unexplained dmp in dryxvcll pressure following an initial rise OlR
- 3. Dryvell pressure > 44 psig
- 2. Drywell pressure response not consistent with LOCA conditions indicating a containment breach
- d. Blreachedl/lBvassed
- 1. Failure of all isolation valves in ANY one line penetrating Primary Containment to close when required NA AND Downstream pathway exists to environment
- 2. Intentional venting per EMG-3200.02 is required NA
- 3. Unisolable primary system leakage outside of dtywell as indicated by exceeding EITIIER of the following in one or more areas requiring a scram:
EMG-3200.11 Max Nomial Temperature NA Olt EMG-3200. 11 Max Nomial Radiation Level
- e. Containment Hvdrogen Concentration
- 1. Containment 112 concentration 2 6%
NA AND Containment 02 concentration 2 5%
F. Emergencv Director Judgment
- 1. ANN' condition in the judgment of the Emergency Director that indicates Loss or Potential loss of the Primary Containment barrier Page 45 of 123 Revision of
Ouster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER DEGRADATION 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 I 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-I for Fuel Clad, RCS and Primary Containment loss and potential loss indicators and bases.
REFERENCE(S)
- 1. NUMARCINESP-007, Revision 2, Section 3.4 & Table 5-F-I NUMARC IC FG I - Table 5-F-I Site Area Emergency DIFFERENCES None Page 46 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER FS1 lNITIATING 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-1, 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 I 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-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-I NUMARC IC FSI - Table 5-F-I Site Area Emergency DIFFERENCES None Page 47 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER FAl INITIATING CONDITION Loss or Potential Loss of EITHER Fuel Clad OR RCS Barriers 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 Fuel Clad barrier OR ANY LOSS or POTENTIAL LOSS of Reactor Coolant System barrier MODE APPLICABILITY I and 2 BASIS 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-I NUMARC IC FSI -Table 5-F-I Alert DIFFERENCES None Page 48 of 123 Revision Of
Ovster Creck Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER FUI 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 MODE APPLICABILITY I and 2 BASIS 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-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-I NUMARC IC FS I - Table 5-F-I Unusual Event DIFFERENCES None Page 49 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER TABLE F-I Barrier Threshold Bases 1.0 FUEL CLAD BARRIER BASES
- a. RPV Water Level Loss:
I. RPVlevel < -30" TAF 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 - WITHI 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.
Potential Loss: 2. RPV'Ie'el < 0" TAFOR CAANNOTbe letermined 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 123 Revision Of
Onster Creek Nuclear Station Annex Exelon Nuclear 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 TAR.
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.
- 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.
REFERENCE(S)
- 1. 2000-GLN-3200.01, Plant Specific Technical Guideline, p. Cl-2
- 2. 2000-BAS-3200.02, EOP Users Guide
- b. Drvwell Radiation Monitoring Loss:
- 1. Containment Hi-Range Radiation Monitoring System (CHRRAS,) > 440 R/ir 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 pLCi/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 Page 51 of 123 Revision Of
luetpr Creek Miloner St-ation Annex Fyplnn Nucleaqr flvgpr
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~~In Nidon Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER 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.
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.
Potential Loss:
None Not Applicable NUMARC IC FC.3
- DIFFERENCES None REFERENCE(S)
- 1. Rad Engineering Calculation No. 2820-99-017
- c. Drvwcll Pressure Not Applicable
- d. Breached / Bvpassed (Primary Coolant Activity Level)
Loss: 1. Coolant iactivity exceeds 300 ;li/gil (DEI)
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.
Potential Loss:
None.
Not Applicable Page 52 of 123 Revision Of
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER NUMARC IC FC.1 DIFFERENCES:
None REFERENCE(S)
- 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 HIdrogen Concentration Not Applicable
- f. Emergency Director Judgment Loss and Potential Loss: 1. Any condition in the judgment of the Emergencys Director that indicates Loss or Potential Loss of the Fuel Clad barrier This threshold addresses any other factors that are to be used by the Emergency Director in determining whether tile 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 NUMARCINESP 007: None.
- 1. 2000-PLN-1300.01, OCNGS Emergency Plan, section 1.1.22 Page 53 of 123 Revision Of
Ovster 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 lowt ered ba 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:
Alone.
Not Applicable NUMARC IC RC.4 DIFFERENCES
- 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.
REFERENCE(S)
- 1. 2000-GLN-3200.01, Plant Specific Technical Guideline, p. Cl-2
- 2. 2000-BAS-3200.02, EOP Users Guide Page 54 of 123 Revision Of
Ovster Creck Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- b. Drwivell Radiation Monitoring Loss: 1. Containment Hi-Range Radiation Monitoring System (CHRRAMS) > 45 R/r l
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.
Not Applicable NUMARC IC RC.3 DIFFERENCES None.
None REFERENCE(S)
- 1. 2000-RAP-3024.01 NSSS Alarm Response Procedures,10-F-4-K
- 2. Rad Engineering Calculation No. 2820-99-017
- c. Drvwell Pressure Loss:
I. Drywell jresstre > 3.0 psig AND indication of a RCS leak inside dlyell 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:
Alone.
Not Applicable Page 55 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER NUMARC IC RC.2 DIFFERENCES
- 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.
REFERENCE(S)
- 1. 2000-BAS-3200.02, EOP User's Guide, p. 2-2
- d. Breached / Bypassed Loss:
- 1. Unisolable AMain Steam Line break outside containment OR
- 2. Unisolable Isolation 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:
- 4. Unisolable primary sjstein leakage outside of drmywell as indicated bar exceeding EITHER of the followring in one or more areas requiring a scram:
EMG-3200.11 AMax Normal Temperature OR EAMG-3200.J11 Alax Normal Radiation Level Unisolable infers that the leak that cannot be isolated from the Control Room.
Page 56 of 123 Revision Of
Ovster Creek Nuclcar Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER 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, 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 drywvell. 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-3200.1 1.
NUMARC IC RC.I DIFFERENCES None REFERENCE(S)
- 1. 2000-GLN-3200.01, Plant Specific Technical Guideline, p. SC-6
- 2. EMG-3200.11, Secondary Containment Control
- e. Containment IHvdroaen Concentration Not Applicable Page 57 of 123 Revision Of
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Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- f.
Emrergencv Director Judgment Loss and Potential Loss: 1. ANY condition in the juldgnment of the Emergency Director that indicates Loss or Potential Loss 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 Directorjudgment that the barrier may be lost or potentially lost.
None REFERENCE(S)
- 1. 2000-PLN-1300.01, OCNGS Emergency Plan, section 1.1.22 Page 58 of 123 Revision Of
Ovster 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 Loss: None l
Not Applicable Potential Loss:
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-3200.01 A RPV Control - No ATWS or EMG-3200.01 B 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).
NUMARC IC PC.4 - NEI 99-01, Rev. 4 DIFFERENCES
- 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 4 of NEI 99-01. Primary Containment Flooding required (entry into SAMG) is now specified in the current NUMARC document.
REFERENCE(S)
Ovster Creek Nuclear Station Annex E~xelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- 5. EMG-3200.02 - Primary Containment Control
- b. Drvwell Radiation Monitoring Loss: None.
Not Applicable Potential Loss:
- 1. Containment Hi-Range Radiation Monitoring System (CHRRMS) >
2.0E+4 R7/r 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%.
None REFERENCE(S)
NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant AccidentsEMG-3200.02, Primary Containment ControlRad Engineering Calculation No. 2820-99-017
- c. Drvwell Pressure Loss:
- 1. Rapid unexplained drop in diyiiell pressure followl!ing an initial rise OR
- 2. Drollwell pressure response not consistent iit/l LOCA conditions indicating a Containment breach Page 60 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER 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.
Potential Loss: 3. Dauntwell pressure > 44psig The threshold pressure is the FSAR drywell design pressure at 2920F.
NUMARC IC PC. I DIFFERENCES None REFERENCE(S)
- 1. OCNGS Technical Specifications section 5.2 Basis (38.1 psig)
- 3. EMG-3200.02, Primary Containment Control
- d. Breached / Bvpassed Loss:
- 1. Failzure ofALL isolation valves in any one line penetrating Primayr Containment to close resultingfirom an isolation actuation signal itwhen required AAND Downstream pathl'air 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. Intentional ventingperEMG-3200.02isreqzirer d
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 Page 61 of 123 Revision Of
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FISSION PRODUCT BARRIER threshold does not apply to venting of Primary Containment as needed to maintain pressure within normal operating limits (1.1-1.3psig)
Loss: 3. Unisolable primary system leakage outside of di ywzell as indicated by exceeding EITHER of the following in one or more areas requiring a scram:
EAIG-3200.11 Aax Normal Temperature OR EAMG-3200. 11 Max Normal Radiation Level 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 NUMARC IC PC.2 DIFFERENCES I. 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.
REFERENCE(S)
- 1. EMG-3200.02, Primary Containment Control Page 62 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- 2. 2000-GLN-3200.01, Plant Specific Technical Guideline, p. SC-6
- 3. EMG-3200.1 1, Secondary Containment Control
- c.
Containment Hydroyen Concentration Loss: Alone Not Applicable Potential Loss: 1. Containment H2 concentration > 6%
AND Dryiwell or tort-s 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.
NUMARC IC PC.I DIFFERENCES None REFERENCE(S)
- 1. EMG-3200.02, Primary Containment Control Page 63 of 123 Revision Of
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Xorw-Nar-a Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- f. Emergencv Director Judgment Loss and Potential Loss: 1. ANY condition in thejtzidgmnient of the Emergency Director that indicates Loss or Potential Loss of the Primary Containment barrier 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, etc.) should also be a factor in Emergency Director judgment that the barrier may be lost or potentially lost.
INUMARC IC PC.6 DIFFERENCES None REFERENCE(S)
- 1. 2000-PLN-1300.01, OCNGS Emergency Plan, section 1.1.22 Page 64 of 123 Revision Of
0Ovster Creek Nuclear Station Annex FExelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER 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 Leakaue/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.
Page 65 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (NI)
SYSTEM MALFUNCTIONS MGI
?IN-ITIATING CONDITION Prolonged Loss of ALL Offsite AC Power AND Prolonged Loss of ALL Onsite AC Power
~EAL THRESHOLD VALUE BOTH 4160V Busses IC and ID 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
- Torus w~ater temperature and RPV pressure exceeds the I-leat Capacity Temperature Limit (Figure F, EMG-3200.02)
,MODE APPLICABILITY I and 2
'BASIS Loss of all AC power compromises all plant safety systems requiring electric power including EGGS, 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, uinder these conditions, fission product barrier monitoring capability may be degraded.
Although it may be difficult to predict w~hen 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:
Page 66 of 123ResinO Revision Of
Ovster Creek Nuclear Station Annex Excion Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (NI)
SYSTEM MALFUNCTI ONS MG1 -Cont'd BtASIS - Co'nt'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 Directorjudgment 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'" TAP equates to water level at TAF.
Core submergence is the mechanism of core cooling wvhereby each fuel element is completely' covered wvith water. Indicated RPV wvater level at or above the top of active fuiel (0" TAP) provides direct confirmation that adequate core cooling exists. Assurance of continued adequate core cooling through core submergence is achieved when RPV wvater level can be maintained at or above TAP. If RPV water lev'el cannot be restored and maintained above the top of active fuel, less desirable means of assuring adequate core cooling must be employ'ed, posing a possible threat to fulel clad barrier integrity.
Even if a procedure instructs deliberately' lowering RPV wvater 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 lev'el indications, in general, the EOP action point "cannot be determined" relates to the condition where the operator has no idea w~here RPV water level is, or cannot determine by' any means available that the RPV water level is above the point wvhere adequate core cooling can be assured.
Torus water temperatures in excess of the Heat Capacity Temperature Limit (H-CTL) 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 wvater 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).
Page 67 of 123Reionf Revision Of
Ovster Crcek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAU Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MG1 - Cont'd BASIS - Cont'd Emergency Buses I C and I D can be powered from non-emergency Buses IA and I B and emergency diesel generators EDG-1 and EDG-2.
Buses IA and IB can be powered from the Auxiliary transformer, and Startup transformers SA and SB.
Bus I B can also be powered by the SBO transformer.
REFERENCE(S)
- 1. EMG-3200.02, Primary Containment Control
- 2. 2000-BAS-3200.02, EOP User's Guide
- 3. 1IOCFR50.63
- 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.
Page 68 of 123 Revision Of
nvetowr t-vooU2Nll~l tli Annoy rynolnn V-1-os Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (1)
SYSTEM MALFUNCTIONS MS1 INITIATING CONDITION Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses LEAL THRESHOLD VALUE BOTH 4160V Busses 1C and ID de-energized for> 15 min.
MODE APPLICABILTTY I 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 wxill cause core uncovering and loss of containment integrity, thus this event can escalate to a General Emergency.
Emergency Buses I C and I D can be powered from non-emergency Buses I A and I B and emergency diesel generators EDG-I and EDG-2.
Buses IA and I B can be powered from the Auxiliary transformer, and Startup transformers SA and SB.
Bus I B can also be powered by the SBO transformer.
REFERENCE(S)
- 1. OCNGS Drawing BR 3000
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MAI 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 iC and ID for> 15 min.
AND EITHER of the 4160V busses IC or ID de-energized for > 15 min.
MODE APPLICABILITY I 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 IA fed from either Startup transformer SA OR Auxiliary transformer OR Bus ID from Bus I B 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 I C and I D can be powered from non-emergency Buses I A and I B and emergency diesel generators EDG-I and EDG-2.
Buses IA and I B can be powered from the Auxiliary transformer and Startup transformers SA and SB.
Bus I B 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 70 of 123 Revision Of
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SYSTEM MALFUNCTIONS MAI -Cont'd NUMARC IC SA5
.DIFFERENCES None Page 71 of 123 Rvso f
Revision Of
Ovster Creek Nuclear Station Annex Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS Exelon Nuclear MA2 IJNITIATING 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 IC and ID de-energized for> 15 min.
MODE APPLICABILITY 3,4 and D 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 I D can be powered from non-emergency Buses IA and I B and emergency diesel generators EDG-I and EDG-2.
Buses IA and I B can be powvered from the Auxiliary transformer and Startup transformers SA and SB.
Bus I B can also be powered by the SBO transformer.
REFERENCE(S)
Ovster Creek Nuclear Station Annex Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS Exelon Nuclear MUI 4INITIATING 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 IC and ID for> 15 min.
,MODE APPLICABILITY 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 powver losses.
Emergency Buses I C and I D can be powered from non-emergency Buses IA and I B and emergency diesel generators EDG-I and EDG-2.
Buses IA and I B can be powered from the Auxiliary transformer and Startup transformers SA and SB.
Bus I B can also be powered by the SBO transformer.
REFERENCE(S)
Ax etnr drnonL Navl-na r -
tnt*nn A nn nv li-sul-NXT-lo 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 I 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
Ovstcr Creek Nuclear Station Annex Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS Exelon Nuclear MU3 INITIATING CONDITION Unplanned Loss of Required DC Power During Cold Shutdown or Refueling Mode > 15 Min.
WEAL 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 iBASIS 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 is 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 LINITIATING CONDITION 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 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 tile 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.
Page 76 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear 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 atTAF.
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 15007F; 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 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 TAR.
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 FICTL 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 77 of 123 Revision Of
Oyster Creek Nuclear Station Annex Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS Exclon Nuclear MG4 - Cont'd DIFFERENCES 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.
Page 78 of 123 Revision Of
Ovster Creek Nuclear Station Annex Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS Exelon Nuclear 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.
Page 79 of 123 Revision Of
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SYSTEM MALFUNCTIONS MS4 - Cont'd REFERENCE(S)
DIFFERENCES None Page 80 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MA4 JNITIATING 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 methods 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.
Page 81 of 123 Revision Of
Ovster Creek Nuclear Station Annex Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS Exelon Nuclear MA4 con't REFERENCE(S)
- 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.
Page 82 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MS5 lNITIATING 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 I 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 -NEI 99-01, Rev. 4 DIFFERENCES
- 1. Implements NEI 99-01 rev. 4. BWR specific criteria. Revision 2 of NUMARC/NESP-007 simply specified loss of [site-specific function] necessary to maintain Hot Shutdown. Revision 4 of NEI 99-01 is specific in defining this condition for BWRs as inability to maintain parameters belowv -leat Capacity Temperature Limit.
Page 83 of 123 Revision Of
Onster Creek Nuclear Station Annex Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS Exelon Nuclear 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 2120F 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 wvell as offsite electrical power transformers and Emergency and SBO Diesel Generators.
REFERENCE(S)
I. OCNGS Technical Specifications definitions section 1.7 NUMARC IC SA3 DIFFERENCES None Page 84 of 123 Revision Of
Ouster Crpek Niulear Station Annex F.Yplnn Nuleagr
. 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
- Sustained power oscillations (30 watts/cm2 LPRM peak to peak)
- Stuck open EMRVs
- ECCS Injections Table M Safety System Annunciators
- ECCS (B, C)
- Containment Isolation (G, HS, J)
- Reactor Scram (G)
- Process Radiation Monitoring (IOF)
Table M Safety Function Indicators
- Reactor Level, Pressure and Power (Panel 4F, 5F, 6F)
- Decay Fleat Removal (Panel I F/2F)
- Containment Safety Functions (Panel I IF, 12XR, 16R)
Page 85 of 123 Revision Of
Ovstcr Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (NI)
SYSTEM MALFUNCTIONS MS6 - Cont'd MODE APPLI'CABlL'ITY I and 2 BASIS This EAL is intended to recognize the inability of the control room staff to monitor thle 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 emergency2 RPV depressurization or ECCS injection, or reactor power oscillations of I10% 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.
iREFERENCE(S)
0%,ster Creek Nuclear Station Annex FEvelnn Nuclea-r Ovstr Crek ucler Satio Anex Fi~n Ni~h'n 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 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-I)
- Plant Process Computer is unavailable Table M Significant Plant Transients
- > 25% thermal power change
- Sustained power oscillations (30 wvatts/cm 2 LPRM peak to peak)
- Stuck open EMRVs
- ECCS Injections Table M Safety System Annunciators
- ECCS (B, C)
- Containment Isolation (G, 11, J)
- Reactor Scram (G)
- Process Radiation Monitoring (IOF)
Page 87 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Oyster Creek Nuclear Station Annex Exelon Nuclear 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)
- Decayl-Heat Removal (PanelIFf/2F)
- Containment Safety Functions (Panel II IF, 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.
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 Lininterruptible 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 wxas selected as a threshold to exclude transient or momentary power losses.
Page 88 of 123ReionO Revision Of
()zbvtcrCreekL Nivelonr Station Annoy Vyoltnn Niirlear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MA6 - Cont'd BASIS - 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 I0% or greater (> 30 watts/cm2 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 NEI 99-01 Rev. 4. This statement does not provide useful assessment criteria to the EAL threshold.
Page 89 of 123 Revision Of
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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 or all of EITHER:
Safety system annunciators (Table M-2)
OR 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 (IOF)
Table M Safety Function Indicators
- Reactor Level, Pressure and Power (Panel 4F, SF, 6F)
- Decay IHleat Removal (Panel I F/2F)
- Containment Safety Functions (Panel I IF, 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.
Page 90 of 123 Revision Of
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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. IO CFR 50.72
- 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 NEI 99-01 Rev. 4. This statement does not provide useful assessment criteria to the EAL threshold.
Page 91 of 123 Revision Of
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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 AEOD/EGO9, "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 tile 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 RGI.
RPV water level instrumentation is referenced to the Top of Active Fuel. 0o TAF equates to water level atTAF.
Page 92 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MS7 - Cont'd BtASIS" 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.
IREFERENCE(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.
Page 93 of 123 Revision Of
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (Ml)
SYSTEM MALFUNCTIONS MU7
,INITIATING CONDITION RCS Leakage EAL' THRESHOLD VALUE Unidentified leakage > 10 gpm OR Identiried leakage > 25 gpm
,MODE A'PPLICABILITY I 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 thle unidentified leakage was selected as it is observable with normal control room indications. Lesser values must generally be determined throughl 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)
I. OCNS Technical Specifications Section 3.3.D NUMARC IC SU5 DIFFERENCES
- 1. The valuiespecified by NUMARC/NESP-007 Rev. 2for unidentifijed leakage (10gpm) 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.
Page 94 of 123ReionO Revision Of
(1vetor CrPA-Niir1Pnr.1qtqf;nn Annoy F.vplnn Nw-1pnr At itpr CrooL¶ NuwIoir t,tmnn Annp'i
-ErPIAn Niip1 p Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MU8 IINITIATING CONDITION Unplanned Loss of All Onsite OR Offsite Communications Capabilities 1EAL THRESHOLD VALUE Loss of all onsite communications (Table M-4) affecting the ability to perform routine operations OR Loss of all offsite communications (Table M-4)
Table M Communications Plant Paging System Conv'entional telephone lines Cell Phones Radio ERF NRC Emergency Notification System (ENS)
Health Physics Netwvork (H-PN)
Bureau of Nuclear Engineering Information Line ED Hotline New Jersey State Police (NJSP Notification Line)
Ocean County Notification Line NJ State ED Hotline Environmental Assessment Direct Line Onsite x
x x
x nsite o ffsite x
x x
x x
x x
x x
x x
x x
x x
MODE APPLICABILITY All Page 95 of 123 R~iinO Revision Of
(Y.'cwfnr ('roont hlnlr 4tZ tinn Ann n 1,.nlfln NW22l 0 0
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-Stl4 I~Sl llEU I'~%X^
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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 an 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
Ch'dotgr C'riok NvvgdoPr 1Rtnfinn Annpy FPxelnn Nitelpir Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (NM)
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 7MODE APPLICABILITY I and 2 BASIS Limiting Conditions of Operation (LCOs) require the plant to be brought to a required shutdown mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specifications requires 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)
- l. OCNGS Technical Specifications
- 2.
NUMARC IC SU2 DIFFERENCES None Page 97 of 123 Revision Or
Oyster Creck Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS 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 tile 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.
Discretionarv This category provides the Shift Manager or Emergency Director the latitude to use discretion in the declaration of emergencies based upon his/her judgment.
Page 98 of 123 Revision Of
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (II)
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 HGI DIFFERENCES None Page 99 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (IH)
HAZARDS AND OTHER CONDITIONS HS1 INITIATING CONDITION Confirmed Security Event in a Vital Area
,EAL 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 1iSI DIFFERENCES None Page 100 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HAl 1NITIATING 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)
MODE 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 intnision 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 IIA4 DIFFERENCES None Page 101 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H1)
HAZARDS AND OTHER CONDITIONS HUI 11NITI'ATING CONDITION Confirmed Security Event That Indicates a Potential Degradation in the Lev'el of Plant Safety' tEA L T'H RE'SHOL ID VA L'UES' I.
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 normnal 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 phy'sical evidence supporting the threat exists, if inform-ation 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 ev'ent 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 uinder 10 CFR 73.71 or 10 CFR 50.72 and will not cause an Unusual Ev'ent to be declared.
This event w~ill be escalated to an Alert based upon a hostile intrusion or act within the Protected Area.
Page 102 of 123ReionO Revision Of
Ov-ster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Tcchnical Basis RECOGNITION CATEGORY (1H)
HAZARDS AND OTHER CONDITIONS HU1 Cont'd REFERENCE(S)
- 1. Letter from Mr. B. A. Boger (NRC) to Ms. Lynette Hendricks (NEI) dated 2/4/02 NUMARC IC HU4 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
- 2. This EAL threshold has been written to conform with IC HU4 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 103 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (I-I)
HAZARDS AND OTHER CONDITIONS HS2 INIlTI'AT ING CONDITION Control Room Evacuation Initiated AND Plant Control CANNOT be Established in
- 15 min.
jELTHRESHOLD "VAILUES Control Room evacuation initiated AND Control of the plant CANNOT be established in
- 15 min. per 2000-ABN-3200.30 "Control Room Evacuation" MNODE APP`LICABILI'TY:
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 ev'ent will be escalated based upon system malfunctions or damage consequences.
REFERENCE(S)
- 1. 2000-AB3N-3200.30 "Control Room Evacuation" NUMARC IC I-IS2 DIFFERENCES None Page 104 of 123ResonO Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (II)
HAZARDS AND OTHER CONDITIONS HA2 INITIATING CONDITION Control Room Evacuation Initiated
.EAL THRESHOLD VALUES Entry into 2000-ABN-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)
- 1. 2000-ABN-3200.30 "Control Room Evacuation" NUMARC IC HA5 DIFFERENCES None Page 105 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H1)
HAZARDS AND OTHER CONDITIONS HA3
~INITIATING C O'ND'IT-ION 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 wvind speeds > I100 mph causing damage to Plant Vital Structures (Table H-I1)
- 3. Report of visible structural damage to ANY Plant Vital Structure (Table 1-1-1) due to natural or destructive phenomena OR
- 4. Vehicle crash damaging or affecting Plant Vital Structure (Table [I-I)
- 5. Abnormal Intake Structure lev'el, as indicated by EITHER:
> 6.0ft. MSL (>4.92 psig on P1-SWS-1[2])
- <- 4.0 ft. MSL (< 0.50 psig on P1-533-1172 or P1-533-1173)
MSL = Mean Sea Level Table H-I Plant Vital Structures Reactor Bldg.
Turbine Bldg.
Control Room Complex Main Transformer/Condensate Transfer Pad Intake Structure
- 1 EDG Vault
Ovester Greek Nvirlear Station Annev Rypinon Nitelenr-Atrt'r Crool Nuw1r 5hitinn Annv Ee1nn Nnrh'nr Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (11)
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 I - 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. I 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 I & 2. This EAL specifies the Plant Vital Structures, which contain systems and functions required for safe shutdown of the plant.
Page 107 of 123 Revision Of
(luctnt-rirnnL, Nivolon v _Qfaf;nn A nnov F.Voinn Nilrld'av Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H1)
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 P1-S WS-I [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 P1-533-1 172 or P1-533-I1173) indicates the possible loss of Emergency! Service Water pumps. Procedures require the unit to be brought to cold shutdowvn.
This event w~ill be escalated to a higher emergency classification based upon damage consequences covered under other various EAL Sections.
REFERENCE(S)
I. 2000-AB3N-3200.38 Station Seismic Event
- 2. FSAR Update Section 3.3.7 (Seismic)
- 3. FSAR Update Section 3.3.1 (High w~inds)
- 4. 2000-ABN-3200.31 High Winds
- 5. 2000-ABN-3200.32 Response to Low Intake Levels
- 6. 2000-AIBN-3200.29 Response to Fire (Plant Vital Structures)
NUMARC IC HAl -NEI 99-0 1, Rev. 4 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 ICHIA 1.6 -No safety'related plant areas are susceptible to turbine failure-generated missiles.
Page 108 of 123Reionf Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (11)
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 P1-SWS-1 [2])
- < -3.0 ft. MSL (< 0.94 psig on P1-533-1172 or P1-533-1173)
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. This event will be escalated to an Alert if the earthquake reaches the level requiring a reactor scram per 2000-ABN-3200.38 Station Seismic Event.
Page 109 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (IH)
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 (2 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 P1-SWS-l [2]) is sufficiently high to require plant shutdown per -ABN-.32 Abnormal Intake Level.. 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 P1-533-1172 or P1-533-1173) 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 110 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (HI)
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 iNUMARC IC HUI DIFFERENCES
Page III of 123 Revision Of
Oldster Creek Nuclear Station Annex E~xclon Nuclear Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Tech nical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HA4
~~~~~~~~~~~~-.
.....,i
., I. <,.............
.....I. - -
1INITIATING 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-I) or affecting one or more Safe Shutdown Systems (Table H-2)
AND Safe Shutdown System operability is required Table H-I 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
- Isolation Condenser
- 4160 Safeguard Busses
- Control Room Ventilation
- SDC (IC& ID)
- 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 112 of 123 Revision Of
Oyster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (II)
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 Safe 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 l-A2 DIFFERENCES None Page 113 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H1)
HAZARDS AND OTHER CONDITIONS HU4
'INITIATING CONDITION Fire Within Protected Area Boundary NOT Extinguished in < 15 min. of Detection TEA TRESHOLD VALUE I.
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 permanent structures or equipment.
Table H-I Plant Vital Structures Reactor Bldg.
Turbine Bldg.
Control Room Complex Main Transformer/Condensate Transfer Pad Intake Structure
- 1 EDG Vault
- 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 114 of 123ReionO Revision Of
nvttrr roL welf
.~tltnA n n A Vlvi~wnn Ntt.l-L-ar Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HU4 - Cont'd lBASIS - 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 of the alarm, in this context, means those actions taken in the Control Room to determine that the control room alarm is not spurious. A verified alarm is assumed to be an indication of a fire unless it is disproved within the 15 minute period by personnel dispatched to the scene. In other words, a personnel report from the scene may be used to disprove a sensor alarm if received within 15 minutes, but shall not be required to verify the alarm.
This event will be escalated to an Alert if the fire damages Safe Shutdown Systems required for the current operating condition.
REFERENCE(S)
- 1. 2000-ABN-3200.29 Response to Fires NU MARC IC IHU2 DIFFERENCES
- 1. Unanticipated explosions from NUMARC IC HU 1.5 has been incorporated into this EAL as a logical precursor to OCNS EAL I-JA4.
Page 115 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (II)
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 oftoxic gases within Plant Vital Structures (Table H-I) in concentrations that will be life threatening to plant personnel.
- 2. Report or detection of flammable gases within Plant Vital Structures (Table H-I) 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
- I EDG Vault
- 2 EDG Vault EDG Fuel Oil Storage Tank MODE APPLICABILITY ALL BASIS 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 116 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HA5 - Cont'd LBASIS -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.
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 117 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (11)
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 MODE 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 118 of 123 Revision Of
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (11)
HAZARDS AND OTHER CONDITIONS HU5 - Cont'd
[REFERENCE(S)
- 1. 2000-ABN-3200.33 Toxic Materials/Flammable Gas Release NUMARC IC H-U3 DIFFERENCES None Page 19 of 123 Revision Of
Oyster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Tcchnical Basis RECOGNITION CATEGORY (II)
HAZARDS AND OTHER CONDITIONS HG6
- INITIATING CONDITION Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency EALTHRESHOLDVALUES
- 1. Actual or imminent core degradation with potential loss of containment.
- 2. Potential uncontrolled radionuclide release, which can reasonably be expected to exceed I Rem TEDE, or 5 Rem CDE Child Thyroid plume exposure levels at the Site Boundary or beyond FNIODE APPLICABILITY'"
ALL BA-SI'S General Emergency - Events are in process or have occurred wvhich involve actual or imminent substantial core degradation or meltingvwith 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 wvith thle 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 wvill be classified uinder Event Category R, "Abnormal Radiological Levels/Effluents".
REFERENCE(S)
'DIFFERENCES None Page 120 of 123 R~iinO Revision Of
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (11)
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
.THRES.H O.LD
.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.
fMODE APPLICABILITY ALL BASIS Site Area Emergencv - 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 wvill be classified under Event Category R, "Abnormal Radiological Levels/Effluents".
REFERENCE(S)
None NUMARC IC IHS3 DIFFERENCES None Page 121 of 123 Revision Of
Ovster 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 TEAL 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.
REFERENCE(S)
None NUMARC IC H-A6 DIFFERENCES None Page 122 of 123 Revision Of
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dn Iu~i' Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (II)
HAZARDS AND OTHER CONDITIONS HU6 J-I~iTATINGC,0 CN`DI-TIO Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Unusual Event EkAL 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 BA S-IS (R-efer'e nces')
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 Ev'ent 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 wvithin acceptable design limits.
Releases of radioactive materials requiring offsite response or monitoring arc not expected to occur at this level unless further degradation of safety systems occurs. Flowev'er, if one does occur, it will be classified uinder Event Category R, "Abnormal Radiological Levels/EfflUents" REFERENCE(S)
None NUMARC IC H-U5 DIFFERENCES None Page 123 of 123ReionO Revision Of