ML033570561
| ML033570561 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/12/2003 |
| From: | Gallagher M AmerGen Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2130-03-20311 | |
| Download: ML033570561 (40) | |
Text
AmerGenSM AmerGen Energy Company, LLC www.exeloncorp.com An Exelon/British Energy Company 200 Exelon Way Kennett Square, PA 19348 10CFR50 Appendix E.IV.B 2130-03-20311 December 12, 2003 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Oyster Creek Generating Station Facility License No. DPR-1 6 Docket No. 50-219
Subject:
Supplement to Oyster Creek Generating Station Implementation of Emergency Action Levels Developed from NUMARC/NESP-007 Methodology
References:
(1) Letter from M. P. Gallagher (AmerGen Energy Company, LLC) to US NRC, dated March 10, 2003 (2)
Email from P. S. Tam (US NRC) to John Hufnagel (AmerGen Energy Company, LLC) dated May 20, 2003 In the Reference 1 letter, AmerGen Energy Company, LLC (AmerGen) proposed a change to the Oyster Creek Generating Station Emergency Action Levels & EAL Technical Bases based on the methodology outlined in NUMARC/NESP-007, "Methodology for Development of Emergency Action Levels, Rev. 2.
In Reference 2, the NRC staff provided specific requests for additional information (RAI) pertaining to the Reference 1 submittal. provides the NRC questions and AmerGen's response to each question. contains the revised EAL Comparison Summary of Differences. Attachment 3 provides the revised Emergency Action Levels & EAL Technical Bases pages.
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.
FOLO)
U.S. Nuclear Regulatory Commission December 12, 2003 Page 2 Lastly, AmerGen hereby requests review and approval of the revised EALs (i.e., those submitted via Reference 1 and the revised EALs attached to this letter) 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
ATTACHMENT I OYSTER CREEK GENERATING STATION Docket No. 50-219 License Nos. DPR-16 Supplement to Oyster Creek Emergency Action Levels & EAL Technical Bases Response to Request for Additional Information Response to NRC Questions on OCGS EAL Submittal
- 1.
The Summary of Differences identifies NUMARC/NESP-007 Rev. #3 as a source of information or guidance for several EALs (see MU6, MA6, MS5). The reviewer is not familiar with this specific document. The NRC staff is cognizant of the following with regard to EAL guidance:
A.
NUREG-0654/FEMA-REP-1 Rev. #1 (NRC endorsed guidance)
B.
NUMARC/NESP-007 Rev. #2 (NRC endorsed guidance)
C.
NEI 97-03 Rev. #3 (Not reviewed or endorsed by NRC, retracted by NEI)
D.
NEI 99-01 Rev. #4 (Under review & endorsement process by NRC)
Are the references in the submittal to NUMARC/NESP-007 Rev.#3 intended to be one of the four the staff is aware of or is this another document that we should have available for review?
Response
The reference within the Oyster Creek submittal will be corrected to indicate NEI 99-01 for reference to MU6, MA6, and MS5.
- 2.
Proposed Initiating Condition RU7 EAL threshold value is in part "Radiation readings >10 times normal at any of the following ISFSI locations:...." For the proposed condition:
A.
'What is the basis for "normal?"
B.
How often is "normal" determined?
C.
How is this value affected by changes in background levels D.
NEI 99-01, Rev. #4, dated August 2000, used a value of >2 times the ISFSI Technical Specification limits as the threshold value. In NUMARC/NESP-007, Rev. #2 basis statement for AU2 indicates the use of 2 R/hr at the face of the storage module, or 1 R/hr at one foot from a damaged module. What is the basis for the selection of "10 times normal?"
E.
What is the basis for changing the current EAL Threshold value?
Response
Radiological surveys are conducted on a routine basis in and around the ISFSI facility (nominally quarterly). These surveys would constitute the unormal" or background radiation levels in the immediate vicinity of the ISFSI. Surveys are conducted once per quarter, and following the addition of any loaded casks to the Horizontal Storage Modules. These post-installation surveys would generate the revised normal" or background radiation levels.
Response to RAI on OCGS EAL Submittal Page 2 of 9 Oyster Creek, in conjunction with the New Jersey Bureau of Nuclear Engineering, established these revised EAL threshold values, in order to establish a conservative limit and level upon which to base an Emergency Declaration. The value of 1Ox normal' is indicative of a significant change in the condition of the ISFSI modules. The changes to the current EAL threshold value have been created in support of requests on the part of the State of New Jersey Bureau of Nuclear Engineering's desire to establish a more conservative EAL threshold value.
- 3.
Regarding Fuel Clad Barrier from Table F-1, the Loss condition and the Potential Loss conditions both contain the attribute "level cannot be determined." When the condition exists that RPV level cannot be determined, which condition for the fuel clad barrier is provided? Is it a "loss of the barrier" or is it a "potential loss" of the barrier? For the following condition, what would be the classification?
"Primary Containment Loss, RCS leak rate >50 gpm and RPV level cannot be determined."
Regarding the RCS Barrier, if RPV level cannot be determined, the RCS barrier is considered "Loss." Since this is the same criteria applied to the Fuel Clad barrier, in this condition, "RPV Level cannot be determined," is it intended that this be a loss of two barriers from this one indicator?
Response
The EAL criteria cannot be determined" will be deleted from the Fuel Clad Barrier loss criteria, such that the Fuel Clad barrier loss will be <-30" TAF. The Fuel Clad potential loss criteria will remain as indicated, RPV level < 0" TAF or cannot be determined which is equivalent to the RCS loss criteria of RPV level
< 0" TAF (not intentionally lowered by procedure) or cannot be determined. This implements the guidance provided in NEI 99-01, under Fuel Clad Barrier Example EALs, with the RCS barrier loss and Fuel Clad potential loss criteria being equivalent.
- 4.
For the Primary Containment Barrier in Table F-1, item 3d.,"Breached /
Bypassed", Loss event 2, intentional venting per EMG-3200.02 is required with drywell pressure >3.0 psig, indicates this threshold does not apply to venting of Primary Containment as needed to maintain pressure below the high drywell pressure setpoint. Is it intended that this exception apply during declared emergency conditions when/if venting containment at pressures below 3 psig?
Does this statement indicate it is appropriate to have a pathway to the environment from the containment during emergency conditions without considering Primary Containment barrier lost when containment pressure is < 3 psig?
Response to RAI on OCGS EAL Submittal Page 3 of 9
Response
The intent for this EAL is to ensure that the Primary Containment barrier is considered lost when EOP conditions have been met that require venting of primary containment (e.g., containment pressure exceeds 3.0 psig). Under emergency conditions unrelated to containment conditions (e.g., system malfunctions, etc.) venting of containment to maintain pressure less than 3.0 psig would not be a containment loss.
- 5.
Regarding MS4, "Auto and manual SCRAM NOT successful," the EAL threshold includes the ARI (alternate rod insertion) function to also fail before meeting the criteria for this classification. Does the ARI have the same scram requirements as RPS for rod insertion? Specifically, does the ARI have to meet the same requirements to shutdown the reactor as RPS in order to be considered functional? Why is it appropriate to include ARI in the same capacity as the RPS scram function when a reactor overpower condition could have already occurred due to the failure of the automatic and manual scram functions?
Response
ARI does not have the same scram requirements as an RPS scram and does not result in the same rod insertion as RPS. ARI was installed to comply with the ATWOS rule, which is to provide a redundant scram capability both manually and automatically but is not required to cause rod insertion fast enough to prevent exceeding core limits. The ARI scram air header vent path is different than RPS and takes substantially longer to initiate and complete rod insertion.
Alternate Rod Injection was included in this EAL threshold in that ARI is considered one of the actions of the Reactor Operator to rapidly insert control rods into the core, and as a manual scram function, it is consistent with the criteria in NEI 99-01, SA2. ARI functions by providing a mechanism to vent the SCRAM air header, separate and diverse from the Reactor Protection System logic. The basis for the associated EALs is failure of the safety system not its actual impact on the fuel. Other EALs on coolant activity, radiation levels and release rates are designed to categorize the appropriate EAL.
Under this EAL, ARI is not included within the RPS capacity, but is included as a diverse, manual SCRAM capability by the Reactor Operator at the reactor console, with existing procedures defining the use of ARI, should RPS and manual scram functions not shutdown the reactor.
Response to RAI on OCGS EAL Submittal Page 4 of 9
- 6.
Regarding MA4, EAL threshold value for EAL 2, the EAL indicates "failure of all manual scram attempts...." Is this intended to mean failure of all methods to initiate a scram or failure of the manual scram (mode switch or pushbuttons) initiating devices to perform after some number of initiations? The basis indicates all means have failed versus all attempts. Please clarify.
If this is intended to be all means of manually scramming the plant, why isn't the condition stated as "any one of the means failing" to be consistent with any other RPS setpoint failure, since the basis (Differences section) indicates this is anticipatory of the failure of the automatic scram signal?
Response
The EAL threshold value will be revised to state that the threshold value is failure of all manual scram methods, rather than attempts. The basis will be revised to indicate that the manual scram methods are those available to the operator at the reactor console (e.g., manual scram pushbuttons, mode switch and ARI). The condition is not stated as any one of the means failing" as that would stipulate that any single manual scram failure would exceed the ALERT threshold.
Condition 1 for the EAL is indicative of a failure of RPS, with a setpoint being exceeded and an automatic scram failure. Condition 2 is indicative of equipment failures that could be anticipatory, however, no setpoint has been exceeded.
- 7.
Regarding MA6 and MU6 (combine 7 and 8?), the EAL threshold value includes a parenthetical Note 1. The reviewer could not find the corresponding note, please clarify or identify location of the note. The Shift Supervisors (SS) opinion was removed from the EAL because it does not provide any useful assessment criteria per the "Summary of differences" document. Why is the SS opinion not removed from the basis section for the same reason?
Response
The parenthetical Note 1 was included on the EAL Matrix, Table OCNS 3-1, the contents of which are included in the 3rd and 4 th sentences of the first paragraph in the EAL basis statement. The reference to the note will be deleted in the EAL threshold value in the EAL basis. The Shift Supervisors opinion criteria was removed from the EAL threshold value, and retained in the basis statement, such that the threshold value was more objective, whereas the basis statement provides additional guidance for decision makers involved in event classification.
- 8.
With regard to MU8, is the Bureau of Nuclear Engineering Information Line, ED Hotline, NJ State ED Hotline or the Environmental Assessment Direct Line manned such that at any time of day or night it is reasonable to expect the phone will be answered in a relatively expeditious manner (i.e., about 15 minutes)? Can each of these lines support exigent offsite notifications?
Response to RAI on OCGS EAL Submittal Page 5 of 9
Response
The Bureau of Nuclear Engineering Information line is an auto-ring circuit between Oyster Creek Emergency Facilities (Control Room, TSC, EOF) and BNE offices in West Trenton. It is not a continuously manned circuit, and there is no assurance that it would be answered within 15 minutes. The ED hotline is between the Control Room, TSC and EOF, and does not communicate directly with the state of NJ; the NJ State ED hotline is an auto ring circuit between NJ OEM offices in Trenton; and the Oyster Creek Control Room, TSC, and EOF. It is not manned 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day and there is no assurance of the phone being answered within 15 minutes. The Environmental Assessment line is between the EOF, TSC and Control Room.
The BNE Information Line and the NJ State ED hotline could be used to conduct off-site communications with NJ state officials, once those officials were informed of the availability of those information circuits. However, these circuits would not be capable of replacing the defined notification circuits or assigned commercial backup phone systems. Additionally, the ED hotline and Environmental Assessment line could be used to communicate with the EOF, which has a separate PBX, and direct commercial lines separate from the OC station circuits.
In the event that these circuits were available, the EAL threshold criteria of loss of all means of offsite communications would not be met.
- 9.
With regard to HG1, EAL threshold value item 2, is it required to have loss of the remote and alternate shutdown panels, or can this be met with a loss of the remote or an alternate shutdown panel in order to impact the capability?
Response
The EAL basis indicates that loss of Remote shutdown capability occurs when the control function of the Remote 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.
Response to RAI on OCGS EAL Submittal Page 6 of 9
- 10.
With regard to HSI, EAL #2, are there other Security related issues such as hostage taking or extortion, that if directed at a plant Vital Area, could rise to the level of a Site Area Emergency? If so, how are they addressed?
Response
The threshold criteria for declaration of a Site Area Emergency is compromise of a Vital Area. If the Protected Area has been compromised, the appropriate classification would be an Alert, in accordance with HA1, and unless the Security issue of hostage taking or extortion actually compromised a Vital Area, the classification would not be escalated.
- 11.
With regard to HUI, EAL #2, what is the basis for the use of <2 hours or imminent? If one supposes a time is not provided as to time of the occurrence, will the event be declared?
Response
This EAL criteria was established in consideration of NRC Interim Compensatory Measures issued after Sept 11, 2001. Should a threat to the station be received, and once evaluated, deemed credible and imminent, per the ICM guidance a NOUE will be declared. In order to provide guidance to decision makers regarding "imminent" the Exelon Threat assessment process uses a 2-hour timeframe. The threat assessment process provides decision-making guidance in evaluating threats to the station, and if specific criteria were met, including a defined time frame for the threat (e.g., imminent), then the Emergency Plan would be implemented. In all cases, actions in accordance with the station Security Plan, Security Contingency procedures and reportablity criteria would be implemented. Specifically, if a report of a threat is received, without a specific time of occurrence, evaluation through the threat assessment process would occur; however, the NOUE would not be declared, as the threat was non-specific as relates to timing.
- 12.
Regarding HA3, there is no EAL to address site-specific indications in the control room. How is NUMARC/NESP-007, EAL HA1-4 addressed for things such as wind speed or judgment?
Response
In the basis discussion for HA3, guidance for decision makers for judgment evaluation is provided, as in visible damage to structures or indications of degraded system response resulting from the phenomena. This is an escalation Response to RAI on OCGS EAL Submittal Page 7 of 9 of HU3, where the destructive or natural phenomena was noted, but specific adverse consequences, such as damage or degraded system response were not identified. Site-specific indications in the Control Room, for wind speed and Intake level, are available.
- 13.
Regarding HA4, the EAL states that to meet the threshold of this EAL, Safe Shutdown System operability is required. The basis further states that the primary concern for this EAL is the magnitude of the fire or explosion and the effects on the safe shutdown systems required for the present operational condition. The focus appears to be on whether the equipment will operate for the current plant mode. Does this EAL require the system to be operating or required to be operating before the declaration is made? Can the system be in a non-operating standby mode and the fire occur requiring classification?
Response
This EAL drives an event classification as a consequence of a fire or explosion impacting the operability of systems required for safe shutdown. The systems required for safe shutdown are contingent upon the plant mode at the time of the event. That is, with the plant already in cold shutdown for instance, assurance of operability of some of the systems identified in Table H-2 is not required. System operability is defined by capability, not operating status. A system in a standby mode, required for operability for the current plant mode, and rendered inoperable as a result of a fire or explosion, would meet the criteria in this EAL for classification of an ALERT.
- 14.
Regarding HU4, EAL 1 includes verification of the fire alarm. However, the basis statement indicates that verification is by operator actions to confirm alarms received in the Control Room. Are the actions to verify alarms limited to those actions that can be accomplished only in the Control Room as indicated in NUMARC/NESP-007?
Response
The basis statement of HU4 will be revised to include the following statement:
"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."
Response to RAI on OCGS EAL Submittal Page 8 of 9
- 15.
Regarding HA5, the basis indicates that areas that require only temporary access and can be supported by the use of respiratory protection should not be considered as meeting this threshold. Would this include such activities as routine log-keeping, conducting surveillances or other normal plant evolutions?
Describe what is included in the phrase "affect safe operation of the plant."
Response
The basis statement for HA5 will be revised to eliminate the statements regarding temporary access. Further review has indicated that such an exception to a consideration of the EAL threshold is inappropriate, and will be removed.
The basis statement for HU5 defines how a gas release is considered to impede normal operations.
'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."
Response to RAI on OCGS EAL Submittal Page 9 of 9 Summary of Changes to the Original Submittal Question I -
Revised the EAL submittal to indicate reference to NEI 99-01 Rev. #4, vice NEI 97-03 Rev. #3 (MU6, MA6, MS5). See pages 58, 82, 88, and 90.
Additionally, revised the EAL submittal to indicate reference to NUMARC/NESP 007 vice NUMARC/NEI 007. (Various EALs) See pages 53, 54, 55, 57, 58, 59, 60, 61, 62, and 82.
Question 3 -
Question 6-Question 7-Revised the EAL submittal to delete the criteria cannot be determined" from the Fuel Clad Barrier loss criteria (Table F-1, Fission Product Barrier Matrix). See pages 9, 43, and 50.
Revised the EAL threshold value to state the failure is of all manual scram methods rather than attempts. Basis to be revised to state what manual scram methods are available to the operator at the reactor console panel (MA4). See page 80.
Parenthetical note reference removed (MA6, MU6). See pages 86 and 89.
Question 14 -
Basis statement revised to define verification of fire alarms to be those actions taken in the Control Room (HU4). See page 114.
Question 15 -
Basis statement revised to eliminate statements related to temporary access (HA5). See page 116.
ATTACHMENT 2 OYSTER CREEK GENERATING STATION Docket No. 50-219 License Nos. DPR-16 Supplement to the Oyster Creek Emergency Action Levels & EAL Technical Bases Revised EAL Comparison Summary of Differences
NUMARC/NESP-007 Summary of Differences.
Rev. 2 to Proposed OCNS Emergency Action Levels Page 1 of 5 NUMARC/
NESP-007 OCNSEAL OCNS EAL Differences/Justiication Example E
IC/EAL AUI.1 RUI None AUI.2 RUI None AUI.3 N/A OCNS does not have telemetered perimeters monitors, therefore this EAL is not required AUIA.4 NIA OCNS does not use automatic initiation of real time dose assessment therefore this EAL is not required AU2.1 RU6 None AU2.2 RU5 None AU2.3 RU7 None AU2.4 RU2 None AA.1 RA)
None AAI.2 RAl None AAI.3 N/A OCNS does not have telemetered perimeters monitors therefore this EAL is not required AAI.4 N/A OCNS does not use automatic initiation of real time dose assessment therefore this EAL is not required AA2.1 RA5 None AA2.2 RA5 None AA2.3 RA6 None AA2.4 RA5 None AA3.1 RA2 None AA3.2 RA2 None AS1.1 RSI None AS1.2 N/A OCNS does not have telemetered perimeters monitors therefore this EAL is not required ASI.3 RS]
None AS1.4 RS]
None AGI.1 RGI None AGI.2 N/A OCNS does not have telemetered perimeters monitors therefore this EAL is not required AGI.3 RGI None AGI.4 RGI None HUI.1 HU3 None HUI.2 HU3 None HUI.3 HU3 None HUI.4 HU3 None Unanticipated explosions covered under HU4 (NUMARC HUI.5 HU3 HU2) since explosions are more logically associated with OCNS threshold HU4
Summary of Differences NUMARC/NESP-007 Rev. 2 to Proposed OCNS Emergency Action Levels Page 2 of 5 NUMARC/
NESP-007 OCNSEAL Example OCNS EAL Differences/Justification IC/EAL HUI.6 HU3 None HU1.7 HU3 Added site specific high and low intake water levels as other conditions appropriate for this IC HU2 HU4 Included unanticipated explosions from HUI.5 HU3.1 HU5 None HU3.2 HU5 None This EAL threshold has been written to conform with IC HU4 HU4.1 HU]
regarding devices as amended and endorsed by the NRC in a letter from Mr. B. A. Boger to Ms. Lynette Hendricks (NEI) dated 2/4/02.
HU4.2 HU]
This wording conforms to the criteria of HU4 as amended and approved by NRC for post-9/1 I security issue resolutions HU5 HU6 None OCNS does not have installed seismic instrumentation to determine if seismic activity is in excess of OBE levels.
Procedure 2000-ABN-3200.38 "Station Seismic Event" HAI HA3 requires the Shift Manager to scram the reactor for conditions in which the seismic activity causes a threat to safe plant operation. This is consistent with earthquakes in excess of OBE levels and consistent with the existing OCNS seismic analysis.
HAI.2 HA3 None HAl.3 HA3 None HAI.4 HA3 None HAI.5 HA3 None HAI.6 N/A No plant safety equipment is potentially impacted by missiles generated by turbine rotating failures at OCNS.
HAI.7 HA3 Added site specific high and low intake water levels as other conditions appropriate for this IC HA2 HA4 None HA3.1 HA5 None HA3.2 HAS None HA4.1 HAI None HA4.2 HAI None HA5 HA2 None HA6 HA6 None HSILI HSI None HS1.2 HSI None HS2 HS2 None HS3 HS6 None
NUMARC/NESP-007 Summary of Differences Rev. 2 to Proposed OCNS Emergency Action Levels Page 3 of 5 NUMARC/
NESP-007 OCNSEAL Differences/Justiication Example IC/EAL HGI.1 HG)
None HG1.2 HG)
None HG2 IIG6 None SUI MU]
None SU2 MU9 None The condition "In the opinion of the Shift Supervisor, the loss of the annunciators or indicators requires increased SU3 MU6 surveillance to safely operate the unit(s)" has not been SU3 MU6 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.
SU4.1 RU4 None The MODE applicability [1,2] is a deviation from NUMARC SU4.2 R U3
[all] in that, the SJAE Radiation Monitor, selected as an 'other 5U.2 RU3 indication' will only be a valid indication of Fuel Clad Degradation mode's [1, 21.
"Pressure boundary leakage" not applicable to OCNS since no SU5 MU7 distinction is made between unidentified or pressure boundary leakage in the OCNS Technical Specifications.
SU6 MU8 None SU7 MU3 None SAl MA 2 None Added "Loss of manual SCRAM capability indicated by failure of ALL manual SCRAM methods to achieve reactor 5A2M44 shutdown" per resolution of NJ BNE concerns and consistency with Hope Creek Station.
SA3 A5 None The condition "In the opinion of the Shift Supervisor, the loss of the annunciators or indicators requires increased SA4 AM6 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]
None SSI MS]
None SS2 MS4 None SS3 MS3 None Implements NEI 99-01, Rev. 4 BWR specific criteria.
SS4 MS5 Revision 2 of NUMARC/NESP-007 simply specified loss of
[site-specific function] necessary to maintain Hot Shutdown.
Summary of Differences NUMARC/NESP-007 Rev. 2 to Proposed OCNS Emergency Action Levels Page 4 of 5 NUMARC/
NESP-007 OCNSEAL OCNS EAL Diffcrcnces/Justification Example IC/EAL Revision 4 of NEI is specific in defining this condition for BWRs as inability to maintain parameters below Heat 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 was not included in this SS5 MS7 EAL.
Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level.
RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
SS6 MS6 None Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level.
SGI MG]
RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
SG2.1 MG4 None Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level.
SG2.2 MG4 RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
FC.I 1.dl None Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine water level.
FC.2 L.a.2 RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
"AND Indication of RCS leak inside drywell" criteria added to IC.3 Lb. I clarify intent and distinguish from loss of containment cooling FC.3 1.b.1 events which can manifiest itself symptomatically similar to RCS leakage which is the intent of the IC.
FC.4 N/A No 'other' fuel clad loss/potential loss indicators identified for FC.4 N/A OCNS FC.5 If I None RC.I 2.c.1 None RC.2 2.a.1 Added the condition "OR CANNOT be determined" consistent
NUMARC/NESP-007 Summary of Differences Rev. 2 to Proposed OCNS Emergency Action Levels Page 5 of 5 NUMARC/
NESP-007 OCNSEAL OCNS EAL Differences/Justification Example IC/EAL 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.
RC.3 2.d. 1/2/3/4 None RC.4 2.b.1 None RC.5 NIA No 'other' RCS loss/potential loss indicators identified for RC.5 N/A OCNS RC.6 2.f1 None PC.I 3.c.1/2/3 -
None 3 e.]
The Revsion 2 NUMARC example EAL prescribes an RPV water level in conjunction with the Maximum Core Uncovery PC.2 3.a.1 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.
PC.3 3.d.1/2/3 None PC.4 3.b.J None PC.5 6IA 3.No
'other' PC loss/potential loss indicators identified for PC.5 N/A OCNS PC.6 3f None
ATTACHMENT 3 OYSTER CREEK GENERATING STATION Docket Nos. 50-219 License No. DPR-16 Supplement to the Oyster Creek Emergency Action Levels & EAL Technical Bases Revised Emergency Action Levels & EAL Technical Bases Pages (For information only)
Pages:
Cover 9 of 122 43 of 122 50 of 122 53 of 122 54 of 122 55 of 122 57 of 122 58 of 122 59 of 122 60 of 122 61 of 122 62 of 122 80 of 122 82 of 122 86 of 122 88 of 122 89 of 122 90 of 122 114 of 122 116 of 122
Ovsrtpr Crpek Nuclear Station Annex.
ExYelnn Ncrlear Oyster Creek Nnelenr St2tinn Annex Exelnn Nm'Ir'nr Oyster Creek Nuclear Station Annex Section 3 Emergency Action Levels (EALs)
EAL Technical Bases Revision Od Prepared By; Operations Support Senices, Inc.
1716 White Pond Lane Waxhaw, NC 28173 (704) 243-0501 www.ossi-nct.com Prepared For Exelon Nuclear 200 Exelon Way Kennett Square, PA 19348 Purchase Order #01042079
I'-Io Url errC#.t1.
A EFeonNucea TABLE OCNS 3-1: Emereency Action Level (AL)
Matrit (Cont'd)
FISSION PRODUCT BARRIER MATRIX (Applicability: Modes 1 & 2 ONLY)
N.
r FISSION PRODUrr HARRIER STATUS F(;'N ENERAL EMERGENCY FSI: SITE AREA EMERGENCY FAI: ALERT Fil: UNUSUAL EVENT Fad Clad - LOSS X
X X-X X
I X
X F
Fuel Clad - POTENTIAL LOSS X
I X
X X
Reactor Coolant Svnem LOSS X
X X
X X
X X
ReactorCoolant System -POTENIIAL LOSS X
I X
X X
X Pnmav Containment-LOSS X
X X
X X
X Pnmaiv Containment-POTENTtAL LOSS I
X I
X
- 1. FUELCLAD BARRIER
.REACTOR COOLANT SYSIEM BARRIER
- 3. PRIMARY CONrAINMENr BARRIER LOSS POTENTALWoss LOSS PO IENtIAL L)S LOSS POIENTIALLOSS 2 RPVleveIlC0 TAF I. RPVleveldcOuTAF(nolintenlionatty Rv Wnter Level I. RF'Vlevel <-30" TA F OR lo~rvrd by procedure) l None None I. Entry into SAMGs as required by OR
~~~~~~~ORe bpoedr)EOPs
- a. ~
~
~
~
~
~
~
AfO teV detetere LIe I.RVlvle A
RO CANNOT be determined CANNOT be determined I. Containment Hi-Range Radiation
- 1. Containment Hi-Range Radistion 1 Containmenl Ili-Range Radiation Monitloring Monitoring System (CHRRMS)
None Monitoring System (CIIRRMS)
None None Monitoting System (CHRRMS)
Stenllerlng
~
~
>440 Rlh
- 45 Whr
> 2.OE+4 R/hr I l Rapid. unexpilaned drop in diyell
- 1.
pryrelniure>3A pg prsture Folowing an initial rise C. Drywell Pressure None None AND None 2.
t Dvel Fmur NIndicaion o RCS leak inide Drywell 2
NoeDywel pressure response not consistent l3 DryelI pressre> 44pslg Indication of RCS leak inside D~~~~~iwntI swit LOCA conditions indicating a containment breads~~~~~~~ontaimwin reac I. Coolant activity ~~~~~
300 psC~~gm (D~~l)
I.
Ulnisolable Main Steamr Line break
- 3. RCS Leakage 3-pt gprn I. Failure of all isolation valves in ANY
- 1. Coolant actiity 1-- 300 pCitlim ueEn) t one line penetrating Primary outsitile cntai u tenOR Containment to close %hen required OR
- 4. Unisolable psimary aysteam leakage AND
- 2. UnisolabelsolationCondenserlube outside otdrywell as indicated by Doirnrean pathway exits to esceeding EITHER of thte following in environment l
one or mon areas requiring a scram l
OR EMO.3200.11I Max Normnal
- 2. Intentional venting per EMO-3200.02 is Temperuurea required viths Dryvell preasure > 3.0
- 3. Uninolabte pritnaeysystem leskage outside of dtywetl un indicated by exceeding EliTlER of te following in one or more area requiring a aeratr EMW-200.1II Max Normal Temperature OR a
EMG-320011 Max Normal Radiation Level a Contalnmenl 1 Cont anIenl Hi concentration 2 6 %
Hlydroen None None None None None AND Conentration Containment O concentration 2 %
f Emergency Director
- 1. ANY condition in the judgment ofthe Emergency Director that indicates Loss or
- 1. ANY condition in the judgment of the Emergency Director that indicates Loss or I. ANY condition in the judgment of the Emergency Director that indicatel Loss Judgment Potential Loss of the Fuel Clad barrier.
Potential Loss of the RCS barrier or Potential Loss of the Primary Containment barrier Page 9 of 122 Revision Od
Ovster reek Ncelear Stahtion Annexy Exeplon Nreqer Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER DEGRADATION Table F-1 Fission Product Barrier Matrix
-1. FUELCLAD BARRIER LOSS POTENTTAL LOSS
- a. RPV Water Level
- b. Drvwell Radiation Monitorine
- 1. Containment Hi-Range Radiation Monitoring System (CHRRMS) > 440 R/hr
- c. Drvwell Pressure NA
- d. Breached / Bypassed (Primary Coolant Activity Level)
- 1. Coolant activity > 300 pCi/gm (DEI)
- e. Containment Hvdrogen Concentration NA
- f. Emergency Director Judgment NA NA NA NA
- 1. ANY condition in the judgment of the Emergency Director that indicates Loss or Potential Loss of the Fuel Clad barrier Page 43 of 122 Revision Od
Oyster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER TABLE F-1 Barrier Threshold Bases 1.0 FUEL CLAD BARRIER BASES
- a. RPV Water Level Loss:
- 1.
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 15000F; the threshold temperature of fuel clad perforation. This water level is utilized to preclude fuel damage when RPV water level is below the top of active fuel (TAF).
The MSCRWL appears in the RPV CONTROL - WITH ATWS procedure when RPV water level is intentionally lowered to reduce reactor power. When RPV water level is deliberately lowered, power instabilities may produce noticeable oscillations in RPV water level and make it difficult to maintain water level exactly at TAF. This level is also used in the RPV CONTROL -NO ATWS procedure when all attempts to restore and maintain RPV water level above TAF have failed.
RPV water level instrumentation is referenced to the Top of Active Fuel. 0" TAF equates to water level at TAF. -30'TAF therefore means that RPV water level is 30" below TAF.
Potential Loss: 2. RPVlevel < 0" TAFOR CANNOTbe determined Core submergence is the mechanism of core cooling whereby each fuel element is completely covered with water. Indicated RPV water level at or above the top of active fuel (0" TAF) provides direct confirmation that adequate core cooling exists. Assurance of continued adequate core cooling through core submergence is achieved when RPV water level can be maintained at or above TAF. If RPV water level cannot be restored and maintained above the top of active fuel, less desirable means of assuring adequate core cooling must be employed, posing a possible threat to fuel clad barrier integrity.
Page 50 of 122 Revision Od
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER Potential Loss:
None.
Not Applicable Differences from NIJMARCJNESP 007: None.
References:
- 1. Rad Engineering Calculation No. 2820-99-012
- 2. Rad Engineering Calculation No. 2820-99-017
- 3. Rad Engineering Calculation No.96-004
- e. Containment Hvdrogen Concentration Not Applicable
- f. Emergency Director Judgment Loss and Potential Loss: 1. Any condition in the judgment of the Emergency 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 the Fuel Clad barrier is lost or potentially lost. The inability to monitor the barrier (e.g., RPV wvater level cannot be determined in the EOPs, etc.) should also be a' factor in Emergency Director judgment that the barrier may be lost or potentially lost.
Differences from NUMARC/NESP 007: None.
- 1. 2000-PLN-1300.01, OCNGS Emergency Plan, section 1.1.22 Page 53 of 122 Revision d
nAvfegr Pv-No lIoi r.Ctfann A nnow FIrlAnAT NvilfIv
.S, -ct t-sa 1
mt.t.t cc.nm ttm.
saa.
Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- 2. RCS BARRIER BASES
- a. RPV Water Level Loss: I. RPVlevel <0" TAF(not intentionally lowered by procedure) OR CANNOT be determined This threshold addresses the potential concern of adequate core cooling implicitly resulting from major failure of plant functions needed for the protection of the public. It is based on the EOP concern that the only mechanism remaining to assure adequate core cooling is steam cooling.
RCS barrier loss RCS.2 is the same as the Fuel Clad barrier potential loss FC.2. Thus, this threshold is both a loss of the RCS barrier and a potential loss of the Fuel Clad barrier, appropriately escalating the emergency class to a Site Area Emergency classification.
RPV water level instrumentation is referenced to the Top of Active Fuel. 0" TAF equates to water level at TAF.
With regard to the various situations involving a loss of RPV water level indications, in general, the EOP action point "cannot be determined" relates to the condition where the operator has no idea where RPV water level is, or cannot determine by any means available that the RPV water level is above the point where adequate core cooling can be assured If the EOPs instruct deliberately lowering RPV water level below the top of active fuel under ATWS conditions, the RCS is not assumed to be lost or challenged as a result.
Potential Loss:
None.
Not Applicable Differences from NUMARC/NESP 007:
- 1. Added the condition "OR CANNOT be determined" consistent with OCNS EOPs for loss of ability to determine RPV water level. RPV water level must be assumed to be below the barrier threshold if RPV water level cannot be determined by any direct or indirect method.
References:
- 1. 2000-GLN-3200.01, Plant Specific Technical Guideline, p. Cl-2
- 2. 2000-BAS-3200.02, EOP Users Guide Page 54 of 122 Revision Od
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- b. Drvwell Radiation Monitoring Loss:
- 1. Containment Hi-Range Radiation Monitoring System (CHRRMS) > 45 R7r 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 Differences from NUMARC/NESP 007: None.
References:
- 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:
- 1. Dryvellpressttre >3.0 psigAND indication of a RCS leak inside drywell I
Drywell pressure in excess of the drywell high pressure scram setpoint is designed to be indicative of a LOCA event. The phrase "and indication of a RCS leak inside drywell" has been added to exclude drywell pressurization events that are not caused by a loss of the RCS barrier (e.g., extended loss of drywell cooling). If this threshold is exceeded, there is a clear indication that a leak of sufficient magnitude exists that prevents drywell pressure stabilization.
Potential Loss:
None.
Not Applicable Differences from NUMARC/NESP 007:
Page 55 of 122 Revision d
Oyster Creek Nuclear Station Annex FExeinn Ncear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER break propagation leading to a significantly larger loss of inventory is possible. RCS leakage is measured by the normal primary system leakage monitoring system and is leakage into the drywell. Under certain conditions, this system may be isolated due to increased drywell pressure caused by the leak. In that case, a "loss" of RCS will be indicated and this "potential loss" of RCS would not impact the classification.
Inventory loss events, such as a stuck open Electro-Mechanical Relief Valve (EMRV),
should not be considered when referring to "RCS leakage" because they are not indications of a break, which could propagate.
Potential loss of RCS based on primary system leakage outside the drywell is determined from secondary containment area temperatures or radiation levels. EOP guidance stipulates that when the secondary containment temperature or radiation maximum normal value has been exceeded for one area, all systems, except those required for EOP actions or fire suppression, be isolated. The reactor may be manually scrammed if the high temperature or radiation level continues to increase and is being caused by an unisolable primary system discharge into the reactor building. Therefore, it is appropriate to direct emergency classification based on elevated secondary containment temperature and radiation levels.
Secondary containment areas and maximum normal operating temperatures and radiation levels are given in EMG-3200.1 1.
Differences from NUMARC/NESP 007: None.
References:
- 1. 2000-GLN-3200.01, Plant Specific Technical Guideline, p. SC-6
- 2. EMG-3200.1 1, Secondary Containment Control
- e. Containment Hvdrogen Concentration Not Applicable
- f.
Emergencv Director Judgment Loss and Potential Loss: 1. ANY condition in the judgment 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 Director judgment that the barrier may be lost or potentially lost.
Differences from NUMARC/NESP 007: None.
References:
- 1. 2000-PLN-1300.01, OCNGS Emergency Plan, section 1.1.22 Page 57 of 122 Revision Od
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y a t at ta ii
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Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- 3. PRIMARY CONTAINMENT BARRIER BASES
- a. RPV Vater Level Loss: None I
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-3200.01A 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).
Differences from NUMARC/NESP 007:
- 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.
References:
- 5. EMG-3200.02 - Primary Containment Control Page 58 of 122 Revision Od
Ovster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- b. Dvell Radiation Monitoring Loss: None.
Not Applicable Potential Loss:
- 1. Containment Hi-Range Radiation Monitoring System (CHRRMS) >
- 2. OE+4 R/hir The CHRRMS reading is a value that indicates significant fuel damage (> 20% clad failures) well in excess of that required for loss of RCS and Fuel Clad barriers. This value assumes 20% clad failures with the subsequent release of RCS volume into the containment. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant. Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, warranting declaration of a General Emergency. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%.
Differences from NUMARC/NESP 007: None.
References:
- 1. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents
- 2. EMG-3200.02, Primary Containment Control Rad Engineering Calculation No. 2820-99-017
- c. Drysell Pressure Loss: 1. Rapid, unexplained drop in drywell pressure following an initial rise OR
- 2. Drywell pressure response not consistent with LOCA conditions indicating a Containment breach Rapid unexplained loss of pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of containment integrity.
Drywell pressure should increase as a result of mass and energy release into containment from a LOCA. In a design-basis LOCA event, drywell pressure is expected to reach 38.1 psig. Thus, drywell pressure not increasing under these conditions indicates a loss of containment integrity.
Page 59 of 122 Revision Od
n-vtr-W-1-l C+oosr.t#;^
Anon-lrnln" NWIl.M.
y atsa A
t t
.. nnr P;.rnlnn N1u. 0 1
0..
Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER Potential Loss: 3. Dryvellpressure > 44psig The threshold pressure is the FSAR drywell design pressure at 292F.
Differences from NUMARC/NESP 007: None.
References:
- 1. OCNGS Technical Specifications section 5.2 Basis (38.1 psig)
- 3. EMG-3200.02, Primary Containment Control
- d. Breached / Bypassed Loss: 1. Failure ofALL isolation valves in any one line penetrating Primary Containment to close resultingfrom an isolation actuation signal when required AND Downstream pathway exists to environment This threshold addresses containment isolation failures allowing a direct flow path to the environment such as failure of both MSIVs to close following a Main Steam Line break or when an isolation is required with open valves downstream to the turbine or to the condenser.
Loss: 2. Intentional ventingperEMG-3200.02 is required iith dryvellpressure
> 3.0 psig Intentional venting of primary containment per the EOPs to the secondary containment and/or the environment is considered a loss of containment. EMG-3200.02, Primary Containment Control, specifies primary containment venting in Step PC/P-3 (for Primary Containment Pressure Limit) and Step PC/G-2 (for detectable hydrogen). This EAL threshold does not apply to venting of Primary Containment as needed to maintain pressure below the high drywell pressure setpoint (3.0 psig).
Page 60 of 122 Revision Od
(Hester Creek Niirlear tatinn Annex Rypinn Nuclear A'tpr Cr''k Nuw1a'ir Stnfmnn Annv E'rplnn NuieInr Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER Loss: 3. Unisolable primary system leakage outside of drywell as indicated by exceeding EITHER of the following in one or more areas requiring a scram:
EMG-3200.11 Max Normal Temperature OR EMG-3200. 11 Max Normal Radiation Level Unisolable infers that the leak that cannot be isolated from the Control Room.
When evaluating this threshold for unisolable primary system leakage, it is appropriate to attempt isolation from the Control Room prior to classification.
Potential loss of RCS based on primary system leakage outside the drywell is determined from EOP area temperatures or radiation levels. EOP guidance stipulates that when the secondary containment temperature or radiation maximum normal value has been exceeded for one area, all systems, except those required for EOP actions or fire suppression, be isolated. The reactor may be manually scrammed if the high temperature or radiation level continues to increase and is being caused by an unisolable primary system discharge into the reactor building. Therefore, it is appropriate to direct emergency classification based on elevated secondary containment temperature and radiation levels.
Secondary containment areas and maximum normal operating temperatures and radiation levels are given in EMG-3200.1 1.
Potential Loss:
None.
Not Applicable Differences from NUMARC/NESP 007:
- 1. Expanded the isolation failure of primary containment isolation valves to include lines without automatic isolation by deleting "resulting from an isolation actuation signal." Failures such as feedwater line break outside primary containment with failure of the check valve to fully close deserve classification as primary containment losses.
References:
- 1. EMG-3200.02, Primary Containment Control
- 2. 2000-GLN-3200.01, Plant Specific Technical Guideline, p. SC-6
- 3. EMG-3200.1 1, Secondary Containment Control Page 61 of 122 Revision d
Oyster Creek Nuclear Station Annex Exclon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (F)
FISSION PRODUCT BARRIER
- e. Containment Hydrozen Concentration Loss: None Not Applicable Potential Loss: 1. Containment H2 concentration > 6%
AND Drywell or torus 02 concentration > 5%
The specified value of 6% hydrogen concentration is the minimum that can support a deflagration. Likewise, the minimum concentration of oxygen required to support a deflagration is 5%. Combustion of hydrogen in the deflagration concentration range creates a traveling flame causing a rapid rise in primary containment pressure. A deflagration may result in a peak primary containment pressure high enough to rupture the primary containment or damage the drywell-to-torus boundary.
This threshold is intended to cover situations in which the hydrogen production is due to the zirconium-water reaction expected in fuel melt sequences. The oxygen component may be achieved through venting the containment or other means are possible. Since the fuel clad must be breached to sustain the a zirconium-water reaction and the RCS must be breached to accumulate high hydrogen concentrations in containment, the threshold is a loss of 2 out of 3 fission product barriers with a potential loss (or actual loss) of the third.
If drywell or torus hydrogen concentration reaches 2.5 %, primary containment flooding is required, directing entry to the SAMGs. The presence of hydrogen concentrations in the deflagration range (6%) is therefore indicative of a severe accident condition.
Differences from NUMARC/NESP 007: None.
References:
- 1. EMG-3200.02, Primary Containment Control
- f. Emergencv Director Judgment Loss and Potential Loss: 1. ANY condition in the judgment ofthe 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 Page 62 of 122 Revision Od
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MA4 INID!ATIN.CONDITI Auto SCRAM NOT successful OR Loss of Manual SCRAM Capability 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
~!OP4PLICABILTyj l
JE Z
Z Z
Z Z
0I=
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 80 of 122 Revision Od
Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MS5 NITA NGCN T
L
E 77 Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown
[ElfHRSOL Torus water temperature and RPV pressure CANNOT be maintained below the Heat Capacity Temperature Limit (Figure F, EMG-3200.02)
~OEAPPLICABILIyj I and 2 1B_
H L
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 Directorjudgment, or fission product barrier degradation.
__f~E!-
- 1. 2000-BAS-3200.02, EOP User's Guide
- 2. EMG-3200.02, Primary Containment Control SS4
- 1. Implements NEI 99-01 rev. 4. BWR specific criteria. Revision 2 of NUMARCINESP-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 below Heat Capacity Temperature Limit.
Page 82 of 122 Revision Od
Ovster Creek Nuclear Station Annex FReAnn Ncle-ar i
SJAWs s 5 _aslS Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MA6 NITJAT1NG 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 gA THRESHOL -
L F
77;.;.
,W Unplanned loss, for > 15 min., of MOST or ALL of EITHER:
Safety system annunciators (Table M-2)
OR Safety function indicators (Table M-3)
AND EITHER:
A significant plant transient is in progress (Table M-1)
OR Plant Process Computer is unavailable Table M Significant Plant Transients Scram
> 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, H, J)
Reactor Scram (G)
Process Radiation Monitoring (IOF)
Page 86 of 122 Revision d
Oyster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MA6 - Cont'd "Significant transient" includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, level/pressure transients such as emergency RPV depressurization or ECCS injection, or reactor power oscillations of 10% or greater (> 30 watts/cm 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.
- 1. None SA4
- 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 88 of 122 Revision Od
nuctobr CrooU llla
.ttn Annex Rvxenn Nrleepar Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MU6 igNITATNG CONDITION
-,",,ffi, 7`
Unplanned Loss of Most or All Safety System Annunciation or Indication in the Control Room > 15 Min.
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, 5F, 6F)
Decay Heat Removal (Panel IF/2F)
Containment Safety Functions (Panel 1 I F, 1 2XR, I 6R)
I and 2 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 89 of 122 Revision d
nyetfor ree Nlrler qfttinn Annex Fxelnn Nrlear flIvcfpr-__.
Crp NI ~
r~~inA.~
~In1urp Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (M)
SYSTEM MALFUNCTIONS MU6 - 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.
- 1. 10 CFR 50.72
- 2. OCNS simulator walkdown SU3
[WDtFlERENCES[:
- 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 90 of 122 Revision Od
(1vcfor CrPk NurPar tatinn AnnP ExYelan Nrlenr Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HU4 - Cont'd WA_
_i" m
L 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.
- 1. 2000-ABN-3200.29 Response to Fires WUM A
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HU2 precRENCES.
HA4.
- 1. Unanticipated explosions from NUMARC IC HUI.5 has been incorporated into this EAL as a logical precursor to OCNS EAL HA4.
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Ovster Creek Nuclear Station Annex Exelon Nuclear Table 3-2: OCNS EAL Technical Basis RECOGNITION CATEGORY (H)
HAZARDS AND OTHER CONDITIONS HA5 - 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.
- 1. 2000-ABN-3200.33 Toxic Materials/Flammable Gas Release HA3 None.
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