NEI 99-06, Response to Request for Additional Information Regarding License Amendment Request 286, Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactor: Difference between revisions

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{{Adams
#REDIRECT [[ML17363A145]]
| number = ML17363A145
| issue date = 12/21/2017
| title = Response to Request for Additional Information Regarding License Amendment Request 286, Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactor
| author name =
| author affiliation = NextEra Energy Point Beach, LLC
| addressee name =
| addressee affiliation = NRC/NRR
| docket = 05000266, 05000301
| license number = DPR-024, DPR-027
| contact person =
| case reference number = NEI 99-06, Rev 6, NRC 2017-0057
| document type = Response to Request for Additional Information (RAI)
| page count = 733
| project =
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:ENCLOSURE NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 286, ADOPTION OF EMERGENCY ACTION LEVEL SCHEME PURSUANT TO NEI 99-01 REVISION 6, "DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR NON-PASSIVE REACTORS" In Reference 1, NextEra Energy Point Beach, LLC (NextEra) submitted a request for an amendment to revise the facility operating licenses for the Point Beach Nuclear Plant (PBNP) Units 1 and 2. Specifically, the proposed change involves revising the Emergency Plan for PBNP to adopt the Nuclear Energy lnstitute's (NEl's) revised Emergency Action Level (EAL) scheme described in NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," which has been endorsed by the NRC. In Reference 2, the NRC staff requested additional information to complete its review of the requested amendment.
Enclosure 1 provides the NextEra response to the NRC staff's request for additional information.
PBNP RAl-1 For proposed EALs RU1, RA 1, RS1, and RG1, explain the purpose of including the f/owrates for the associated ventilation alignments, and describe what a decision maker would do if the flowrates could not be determined to be as listed. NextEra Response RAl-1 The supporting calculation assumes flowrates for the different ventilation system alignments and these flowrates are used in the calculation of threshold values for the different alignments.
The flowrates were carried over from the calculation results tables to the EALs. The decision makers use only the radiation monitor reading for the applicable system alignment to make the classification.
Therefore, the flowrates have been removed from the proposed EALs. The updated EALs are provided in the Attachments to this submittal.
PBNP RAl-2 Please address the following for proposed EALs RA 1, RS1, and RG1: a. Explain why the proposed radiation monitor setpoints are significantly higher than the current radiation monitor setpoints.
: b. It was not clear to the NRG staff if PBNP could accurately assess offsite dose based on steam line radiation monitors [1 (2)RE-231 and 1 (2)RE-232]
for the range of steam generator pressures that may exist following a wide range of events. Additionally, the atmospheric steam dump or steam generator safety valve may not be fully open, which for either of these conditions, could result in an unnecessary declaration of a General Emergency classification.
Explain how the steam line radiation monitors 1 (2)RE-231 and 1 (2)RE-232 can provide an accurate indication of dose based on setpoint values alone or revise accordingly.
Page 1 of 12 NextEra Response RAl-2a The previous EAL threshold values for RA 1, RS 1, and RG 1 are not directly comparable to those currently in use and those proposed in this submittal.
A standardized methodology for completion of these calculations was not available until NRC endorsement of NEI 99-01, Revision 5, which contained a new Appendix A, Basis for Radiological Effluent EALs. The earlier PBNP EAL threshold values had been determined using site dose projection software in use at that time and were based on assumed meteorology and release parameters.
Example input parameters previously used in calculation of threshold values were: Wind speed of 10 mph with stability class D Release duration of 4 hours used Source term was estimated from site accident dose projection software based on: o Release starting at 4 hours after reactor shutdown o No containment sprays o For SGTRs, water level below the U-tubes (reduced iodine retention fraction)
In contrast, the proposed EAL threshold values have been calculated using site-specific parameters consistent with the updated ODCM, and the latest endorsed guidance of NEI 99-01; notably:
* x/Q values are taken from the PBNP ODCM directly rather than calculated from meteorological inputs
* Source term nuclide specific activity fractions are based on annual activity release data consistent with the PBNP ODCM
* Dose conversion factors (DCFs) for early phase of nuclear accidents are taken from EPA 400-R-92-001
* Release duration of 1 hour used These updated calculated values have been adopted in the current PBNP EALs, which were issued on March 8, 2017 as Revision 30 to the PBNP Emergency Plan, Appendix B, Emergency Classification.
This revision was submitted to the NRC by NextEra under letter number NRC 2017-0015, dated March 17, 2017, titled "Report of Changes to Emergency Plan." In this submittal, NextEra stated that: EP Appendix B was revised due to interdependent updates to the Offsite Dose Calculation Manual and Calculation 2013-0018 (EC 288117) resulting in a revision to the emergency action levels and also includes editorial changes. These changes were evaluated in accordance with 10 CFR 50.54(q)(3) and it was determined that these changes did not reduce the effectiveness of the PBNP Emergency Plan. Page 2 of 12 NextEra Response RAI-2b The 1 (2)RE-231 and 1 (2)RE-232 radiation monitors are designed to provide post-accident effluent monitoring in compliance with Regulatory Guide 1.97, "Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants." These instruments monitor the main steam lines to determine if there are any significant radiological releases from this potential effluent pathway. The calculation of a single threshold value for Initiating Conditions (IC) RA1, RS1, or RG1 does require the use of multiple assumptions (e.g., primary to secondary leak rate, effluent pathway (e.g., safety valve(s) or atmospheric relief), and release rate) all of which could result in an incorrect assessment of these I Cs in several credible scenarios.
Additionally, these instruments (1 (2)RE-231 and 1 (2)RE-232) do not have the capability to quantify an effluent release rate and do not have an established ODCM methodology for determining effluent monitor alarm setpoints.
Because of these factors, and in consideration of the variable system flow rates that may exist during a steam generator tube rupture (SGTR) event, the use of a single calculated value for these instruments to determine if an EAL threshold is exceeded cannot be reliably used as the sole criteria for the declaration of I Cs RA 1, RS1, or RG1. Therefore, PBNP proposes to implement an EAL scheme that will not use the 1 (2)RE-231 and 1 (2)RE-232 radiation monitor threshold values as declaration criteria for ICs RA1, RS1, or RG1. The revise proposed EALs are shown in the Attachments to this submittal.
While this proposed change is a reduction of the current means available to assess whether the initiating conditions of RA 1, RS1, or RG1 exist, multiple and diverse means of assessment do remain available to EAL decision-makers in the form of alternate EALs within these ICs. In the event of a radiological release from the steam line pathway, EAL decision-makers can determine that conditions exist to warrant a declaration of I Cs RA 1, RS1, or RG1 by means of real-time dose assessment using actual steam flow and meteorology at the time of the event, or by means of field survey results that exceed pre-determined threshold values (alternate EALs within the same IC). PBNP maintains a full dose assessment capability (Unified RASCAL Interface (URI)) that uses the 1 (2)RE-231 and 1 (2)RE-232 radiation monitors, along with actual steam line flow rates, to perform off-site dose projections.
At PBNP, dose assessment is an on-shift capability performed by a minimum staffing assignee trained to fulfill the dose assessment function.
The proposed change to not use the 1 (2)RE-231 and 1 (2)RE-232 radiation monitor instrument values as declaration criteria for ICs RA1, RS1, or RG1 does not detract from the guidance in NEI 99-01, the requirements of 10 CFR 50.47(b)(4), or the standards in Appendix E to 10 CFR 50. Additionally, the proposed change does not represent an EAL scheme change. PBNP RAI-3 PBNP proposed EAL RA3. 1 includes the Central Alarm Station (GAS) and the Secondary Alarm Station (SAS) as threshold criteria.
Typically only one of these areas is required.
If the GAS and SAS can both provide access to areas required to assure safe plant operations, explain why EAL does not provide an "AND" logic to the GAS and SAS, or select the primary station as the threshold value as provided in accordance with endorsed guidance.
NextEra Response RAI-3 PBNP added the "AND" logic to the CAS and SAS listing in EAL RA3.1 to clarify that each location is capable of providing access to areas required to assure safe plant operations.
Therefore, loss of both CAS AND SAS would be required to meet the threshold intent of an actual or potential substantial degradation of the level of safety of the plant. The revised proposed EAL is shown in the Attachments to this submittal.
Page 3 of 12 PBNP RAI-4 Tables in proposed EALs RA3 and HA5 contain areas that do not appear to require entry to either maintain normal operation or to shut down and coo/down the plant (e.g., plants typically can transition from Mode 1 to Mode 3 without being req(Jired to enter the turbine building.)
Additionally, plants can typically open the reactor trip breakers from the control room. Please verify all the listed rooms or areas are restricted to only those areas that contain equipment needed for safe operation or safe shutdown I cool-down, or revise accordingly consistent with endorsed guidance.
NextEra Response RAI-4 Following receipt of the RAI, a step-by-step analysis was performed that reviewed all P8NP procedures that contained critical operational activities conducted outside of the Control Room used to implement the normal at-power procedure and the procedures used to shut down and cooldown the plant. This analysis considered the "rooms or areas that contain equipment which require a manual/local action" as noted in the NEI 99-01, Revision 6 developer notes. The analysis included the following P8NP procedures:
* OP 3A: Power operation to Hot Standby
* OP 38: Reactor Shutdown
* OP 3C: Hot Standby to Cold Shutdown
* OP 7A: Placing RHR System in operation
* OP 50: Part 3: Preparation for Chemically Degassing the RCS
* OP 50: Part 4: Degassing the RCS using the PZR and Letdown Gas Stripper This analysis did not include rooms or areas in which actions of a contingent or emergency nature would be performed. (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations).
This analysis also did not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
The resulting list of rooms or areas that contain equipment needed for safe operation or safe shutdown / cool-down now reads: SAFE OPS, SID, CID AREAS Area/Building MOD~ U1 VCT Area 3/4/5 U2 VCT Area 3 /4 / 5 U1 Primary Sample area 3 U2 Primary Sample area 3 CCWHXRoom 4/5 C-59 area 3/4/5 Pipeway 2, 8 ft. elev. 3/4 Pipeway 3, 8 ft. elev. 3/4 1 /28-32 MCC Area 4 This listing has been updated in both IC/EAL RA3 and HAS, as shown in the Attachments to this submittal, and a summary of the step-by-step analysis is provided in the supporting technical documentation of Attachment
: 4. Page 4 of 12 PBNP RAl-5 Proposed EALs CU1, CA1, CS1, and CG1, affecting RCS inventory, include only Containment Sump "A" level rise as an indication of RCS leakage. Previously approved PBNP EALs also included Waste Holdup Tank level rise as an indication of RCS leakage. Please explain the basis for deleting Waste Holdup Tank level rise as an indication of RCS leakage, or revise to include sump and or tank indications that would be indicative of a RCS leak. NextEra Response RAl-5 Waste Holdup Tank level rise as an indication of RCS inventory loss has been returned to proposed EALs CU1, CA1, CS1, and CG1. The revised proposed EALs are shown in the Attachment to this submittal.
PBNP RAl-6 The basis for proposed EALs CU2, SU1, and SA 1 include the following:
Unit 1 (2) offsite power sources include:
* 345 KVAC system supplying power to the 13.8 KVAC system and the 1 (2)X04 transformer
* cross-tying with the opposite unit power supply
* Power to the 1 (2)X -02 Auxiliary transformer through the 19 KV AC system and the 1 (2)X-01 main step-up transformer The capability to cross-tie AC power takes credit for the redundant power source for this IC [initiating condition].
The inability to implement the cross-tie within 15 minutes warrants declaring a UE [Notification of Unusual Event]. However, the staff could not determine which of the above sources of power could not be aligned within 15 minutes. a. Provide an explanation as to the meaning of "cross-tying with the opposite unit power supply," include how cross-tying is supported by abnormal operating procedures (AOPs) or emergency operating procedures (EOPs). b. Explain what is meant by "[t]he inability to implement the cross-tie within 15 minutes warrants declaring a UE." Address why it appears that declaring a UE for not being able to perform the cross tie within 15 minutes would be appropriate if you did not have one power supply already available to an emergency bus, as this condition would warrant the declaration of an Alert or Site Area Emergency classification under EAL CA 1 or SS1, respectively.
: c. For EAL SA 1, correct the UE reference in the following SA 1 Basis statement (for an Alert declaration): "The inability to implement the cross-tie within 15 minutes warrants declaring a UE." Page 5 of 12 NextEra Response RAl-6 The listing of power sources has been updated in EALs SA 1 and CU2 to more clearly reflect the readily available normal power sources used in plant procedures.
Since EAL SU1 only deals with offsite power capability, no listing of normal power sources is provided in the Basis for that EAL. a. Cross-tie of the opposite unit's power supply is controlled by procedure ECA-0.0, LOSS OF ALL AC POWER, steps 23-25 for Train A and steps 41-43 for Train B. For loss of a single emergency bus, procedure AOP-19A, TRAIN "A" SAFEGUARDS BUS RESTORATION (Unit 1, Train A example), steps 16-17 control use of the opposite unit's power to energize the emergency bus. Both procedures are provided as additions to Attachment 4 -Updated Supporting Technical Information as validation documents V36 and V37. b. This statement has been removed from the proposed Basis of the applicable EALs. c. This statement has been removed from the proposed Basis of EAL SA 1. The revised proposed EALs are shown in the Attachments to this submittal.
PBNP RAl-7 The basis for proposed EALs CA2, SS1, and SG1 include the following:
* Unit 1 (2) offsite power sources include: 345 KVAC 1(2)X-03 through the 13.8 KVAC system to the 1(2)X04 transformer 345 KVAC through the 19 KVAC system to the aux transformer 1(2)X-02
* Unit 1 (2) onsite power sources consist of: emergency diesel generators gas turbine generator unit main turbine generator power supplied from the opposite unit Considering that threshold values EALs CA2, SS1, and SG1 are a loss of all offsite and onsite AC power to the emergency buses, the above list of power sources is not required.
Additionally, the inclusion of this table in the Basis discussion for EALs CA2, SS1, and SG1 could imply that a decision maker would not potentially make a declaration for a loss of all AC power because a power supply that is not included in the Basis discussion is providing power to the emergency bus. Please explain why the list of power sources is provided in the Basis discussions for EALs CA2, SS1, and SG1, or revise according to remove list. Page 6 of 12 NextEra Response RAl-7 The additional information in the Basis for these proposed EALs CA2, SS1, and SG1 has been removed to prevent potential confusion to the decision makers. The following additional statement has been added to the EALs CA2, SS1, SG1, and SG2 (SGS in NEI 99-01, Revision 6) Basis discussions to clarify to the decision makers that the use of other than normal power supplies can be credited in the evaluation of these EALs: If mitigative strategies establish emergency power to any bus listed in the EAL, the EAL threshold for this Initiating Condition is not met. The revised proposed EALs are shown in the Attachments to this submittal.
PBNP RAl-8 Description of communications systems appears to be inconsistent with that described in the PBNP Emergency Plan. a. Explain the difference between a commercial phone system, general telephone lines and a private branch exchange (PBX). b. Verify that communications systems listed in EALs are consistent with those provided in the PBNP Emergency Plan and emergency plan implementing procedures.
NextEra Response RAl-8a The PBNP onsite private branch exchange (PBX) system is located on the Unit 1 Turbine Deck area and provides telephone extensions to the Control Room and onsite emergency facilities.
The site PBX is normally powered from a plant lighting circuit and has a battery backup power supply in the event of a loss of power to the lighting circuit. The commercial telephone lines are external lines not powered or fully switched by the site PBX. These lines are also not supplied with backup power from the PBX battery backup power supply. Links between the Commercial lines and the PBX are provided to allow interoperability and limited fail-over protection (e.g., eight in-plant extensions automatically take over eight Mishicot, Wisconsin commercial lines upon loss of all in-plant PBX system power). The term "general telephone lines" was previously used as a generic term for any desktop telephone which was not part of a dedicated network such as the Two-Digit Dial Select circuit (aka NARS Phone) or FTS2000. This term is no longer used at PBNP and has been removed from the proposed EALs. The revised proposed EALs are shown in the Attachments to this submittal.
NextEra Response RAl-8b PBNP staff has reviewed the matrix of emergency response communications provided in the PBNP Emergency Plan, Figure 7-1, as well as applicable implementing procedures and have updated the listing provided in EALs CU5 and SU7 to be consistent.
The revised proposed EALs are shown in the Attachments to this submittal.
Page 7 of 12 PBNP RAl-9 The proposed EALs CS1 and CG1 do not appear to have been developed in accordance with endorsed guidance.
The current PBNP EALs CS1 a*nd CG1 refer to reactor vessel level indications of 0% on Ll-447/Ll-447A and 20 feet Reactor Vessel Level Indicating System (RVLIS) Narrow Range; however, these levels were not included in the proposed EAL scheme, Additionally, no level was provided that was 6 inches below the bottom ID of the reactor coolant system loop. a. Provide further justification for the removal of an EAL that relies on the RVLIS, or revise accordingly. (Note: NE/ 99-01, Revision 6, developer notes provide guidance for indication that is "approximately the top of active fuel.'? b. Provide Reactor Coolant System (RCS) level indication available near the bottom ID of the RCS loop, or explain why this is not addressed.
: c. Explain why an indication that is normally available while in shutdown cooling was not used to provide a site-specific RCS level for CS1 or revise accordingly.
NextEra Response RAl-9a, b, c In the PWR Developer notes provided by NEI 99-01, Revision 6 for these EALS, the following guidance is provided:
If the availability of on-scale level indication is such that this level value can be determined during some shutdown modes or conditions, but not others, then specify the mode-dependent and/or configuration states during which the level indication is
* applicable.
If the design and operation of water level instrumentation is such that this level value cannot be determined at any time during Cold Shutdown or Refueling modes, then do not include EAL #2(1) (classification will be accomplished in accordance with EAL #3(2)). (emphasis added) The Reactor Vessel Level Indicating System (RVLIS) Narrow Range is only calibrated for use during hot conditions.
In the Cold Shutdown or Refueling modes, this instrument is only intended to be used for trend information and is not used for definitive Reactor Vessel level indication.
Additionally, level indicators Ll-447/Ll-447A have a limited range, with 0% being the bottom of the Hot leg nozzle and a reading of 10% defined as the lowest usable on-scale reading. For these reasons, neither instrument is an appropriate indicator for use in these EALs. Therefore, the guidance cited above was followed and these EALs of each IC were not used in the PBNP EAL scheme. Note that the specific Developer guidance provided by NEI 99-01, Revision 6 as quoted above was not provided by the previously NRG-endorsed revisions of NEI 99-01; therefore, previous EALs based on that prior guidance included additional indications which were not well suited for this application.
Page 8 of 12 PBNP RAl-10 For the proposed fuel clad and RCS fission product barriers, RED entry conditions Critical Safety Function Status Tree (CSFST) for the heat sink are used as a threshold for a potential loss of either of these barriers.
However, endorsed guidance states: In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Please explain why the endorsed guidance concerning making classifications for heat sink conditions when operators intentionally reduce heat removal capability, in accordance with EOPs, is not included in the fission product barrier thresholds as this could result in an inaccurate EAL declaration, or revise accordingly.
NextEra Response RAl-10 The guidance cited in the RAI was left out of the Basis document for the two barrier Potential Loss thresholds during reformatting. This guidance has been added to the PBNP Fission Product Barrier Basis for both Fuel Clad Barrier Potential Loss 1.8 and RCS Barrier Potential Loss 1.A. The revised proposed EAL Basis is shown in the Attachments to this submittal.
PBNP RAl-11 The current PBNP Fission Product Barrier (FPB) EAL FC4 Potential Loss includes reactor vessel level indications by RVLIS. Please explain why these indications are not used in proposed FPB EAL FC2 NextEra Response RAl-11 RVLIS was previously only used in the PBNP Fission Product Barrier thresholds as a site-specific piece of the Core Cooling Critical Safety Function Status (CSFS) Orange Path logic. In NEI 99-01 Revision 6, "Basis Information for PWR EAL Fission Product Barrier Table 9-F-3," the Developer Notes for Threshold Parameters and Values state that a plant can use the CSFS either in lieu of plant parameters or in addition to listing the component pieces of the CSFS logic. This guidance was not previously provided in prior revisions of NEI 99-01. Using this new guidance, PBNP only used the CSFS as the method for the decision maker to quickly determine the EAL status since the operators continually monitor CSFS. Excerpt from NEI 99-01, Revision 6 showing this guidance is provided below: The CSFST thresholds may be addressed in one of 3 ways: 1) Not incorporated; thresholds will use parameters and values as discussed in the Developer Notes. 2) Incorporated along with parameter and value thresholds (e.g., a fuel clad loss would have 2 thresholds such as "CETs > 1200°F" and "Core Cooling Red entry conditions met". 3) Used in lieu of parameters and values for all thresholds.
Option 3 of the guidance cited above was used by PBNP in the development of the proposed EALs; therefore, the individual components of the Core Cooling CSFS logic tree are not listed separately.
Page 9 of 12 PBNP RAl-12 Proposed EAL HU4.2 excludes the Containment from required verification of existence of a fire within 30 minutes of a single alarm in Modes 1 and 2. Please provide a justification that supports the HU4 note that excludes the Containment from consideration.
NextEra Response RAl-12 This exclusion was identified in the Enclosure to PBNP License Amendment Request 286 as Proposed Deviation
#2 from the NRG-endorsed guidance provided in NEI 99-01, Revision 6. The explanation and justification previously provided is restated and editorially updated below. Proposed Deviation
#2 NextEra proposes to make an exception in EAL HU4.2 to exclude from classification a single fire alarm in Containment, during Modes 1 and 2 only. Accessing Containment within 30 minutes to verify the status of a single fire alarm presents a personal safety risk, particularly in these Modes when Containment integrity is set and personnel safety concerns would preclude entry into certain areas of the Containment structure.
There are also areas within Containment where fire detectors are located that would be inaccessible during these Modes due to elevated radiation levels. Therefore, verification of a single Containment fire alarm that may or may not be spurious would involve elevated risks (both industrial and*radiation safety) associated with an emergency entry of Containment in Modes 1 and 2. Therefore, NextEra proposes to make EAL HU4.2 applicable to a single fire alarm in Containment in all plant conditions other than Modes 1 or 2. Based on prior industry and PBNP experience, if Containment were to be included in EAL HU4.2 during Modes 1 and 2, a high potential exists for an unwarranted number of Unusual Event emergency classifications based on single spurious fire alarms. Additionally, at PBNP the Containment buildings are designated as NFPA-805 Low Safety Significant (LSS) fire areas, each containing 40 Photoelectric smoke detectors across three levels. The detectors and Fire Alarm Control Panels (FACP) have "smart" technology incorporated into all FACPs. Panels FACP-007 and FACP-008 have the Containment smoke detectors attached to them, however all FACPs onsite show all alarming detectors, providing the operating crew information on the alarming indicator to quickly ascertain the location of the alarming indicator.
If a detector actuates, the actuation drives an alarm in the Main Control Room. All FACPs display the information from the alarming detector as well as the Control Room "Fireworks" panel. There are four Containment Fan Cooler (CFC) units located throughout the Containment building.
Each CFC has one Containment Accident Fan and one Containment Cooling fan delivering a combined 58,000 CFM. Three of the four CFCs are operating at any given time to cool the Containment.
The units draw return air into each end of the unit and discharge into a common header. This constant flow of air (approximately 174,000 CFM) would draw any smoke towards the cooling units past the installed detectors, thus initiating multiple smoke detector alarms. Actuation of more than one smoke detector on the FACP system is the most reliable indication of an actual fire because of high volumetric air flow throughout the Containment building increasing the probability of any actual fire being sensed by multiple detectors.
Due to construction of the intermediate floors and multiple openings in the floors, it can be expected that smoke would migrate throughout Containment in a very short period and that multiple (2 or more) smoke detectors would alarm. Basing emergency classifications on receiving more than one (>1) smoke detector actuation on the FACP system is therefore the most reliable indication of a valid ~larm and accurately meets the fnitiating Condition of HU4, "FIRE potentially degrading the level of safety of the plant." Page 10 of 12 The structure of the proposed deviation for HU4 IC/EAL is modelled after Seabrook Station's adoption of NEI 99-01 Revision 6 containing a similar exception, which was approved by the NRC in a Safety Evaluation dated February 10, 2017 (ML 16358A411
). Consistent with the guidance in Regulatory Issue Summary (RIS) 2003-18, Supplement 2, Use of Nuclear Energy Institute (NEI) 99-01, "Methodology for Development of Emergency Action Levels," Revision 4, dated January 2003, it is reasonable to conclude that the changes proposed to EAL HU4.2 would be considered a Deviation from the formally endorsed guidance of NEI 99-01, Revision 6. PBNP RAl-13 The basis for proposed EAL HS6 states that the Operations Manual assumes the earliest operator action will be taken from a remote shutdown location is 30 minutes. The endorsed guidance provides a typical time of 15 minutes or a time based on a site-specific fire response analysis.
Please provide a site-specific analysis that supports a response time of 30 minutes, or revise accordingly in accordance with endorsed guidance.
NextEra Response RAl-13 PBNP choses to return to use of the default 15 minute time frame provided in the Developer Notes of the endorsed guidance.
The revised proposed EAL is shown in the Attachments to this submittal.
PBNP RAl-14 EAL SU4 does not indicate whether or not the Technical Specification allowable limits, as described by the RU3 IC, include completing required actions within the completion times as provided by the Technical Specifications.
Please clarify whether or not Technical Specification completion times should be considered when assessing RU3. NextEra Response RAl-14 PBNP understands that this RAI contains a typographic error and was intended to discuss IC SU3 of NEI 99-01 Revision 6 (renumbered to SU4 at PBNP), and PBNP answers accordingly.
PBNP has revised the EAL statement as follows to ensure allowed completion times are correctly taken into account by the decision makers when assessing the EAL: SU4.2 Sample analysis indicates that a RCS Specific Activity value is greater than an allowable limit specified in Technical Specifications as indicated by ANY of the following conditions:
: a. Dose Equivalent 1-131 greater than 50 µCi/gm OR b. Dose Equivalent
/-131 greater than 0.5 µCi/gm but less than or equal to 50 µCi/gm for greater than 48 hours OR c. Dose Equivalent Xe-133 greater than 300 µCi/gm for greater than 48 hours The revised proposed EAL is shown in the Attachments to this submittal.
Page 11 of 12 PBNP RAl-15 The final two paragraphs in Basis for proposed EAL SG1 appear to provide direction "to give the Emergency Director a reasonable idea of how quickly the need to declare a General Emergency." In addition to inappropriately including potential procedural direction in the Basis discussion, this direction is not consistent with endorsed guidance.
Please remove the last two paragraphs from the SG1 basis discussion or provide justification that the wording, which is not consistent with endorsed guidance, could not influence and Emergency Director from prematurely declaring a General Emergency.
NextEra Response RAl-15 The last two paragraphs from the SG1 basis discussion have been removed. The revised proposed EAL is shown in the Attachments to this submittal.
Page 12 of 12
* ATTACHMENT 1 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 286, ADOPTION OF EMERGENCY ACTION LEVEL SCHEME PURSUANT TO NEI 99-01 REVISION 6, "DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR NON-PASSIVE REACTORS" UPDATED REDLINE MARKUP OF NEI 99-01 REVISION 6 289 pages follow NEI 99 0:1. [Revisien
&) Development of Emergency Action Levels for Non-Passive Reactors Nevember 2012 
[THIS Pl\GE IS LEFT BLl\NK INTENTIONA.LLY]
NEI 99 01 [Re~isien
&) Nuclear Energy Institute Point Beach Emergency Action Levels Bases Document TBD, 2018 111 
 
(THIS PAGE IS LEFT BLANK INTENTIONALLY]
V Ne"lember 2012 Nt1el-ee1r E,,agy Institute, 177G l Street N. IV., Suite 4FHJ , We1shirtgten D. C. (2(}2. 739. 8{){)~
ACKNOWLEDGMENTS This document was prepared by the Nu c lear Energy Institute (NEI) Emergency Action Level (EAL) Task Force. NEI Chairpersee:
David Young Preparatiee Team Larry Baker E>celon Nuclear/Corp o rate Craig Banner PS E G l'J"uclear/Salem and Hope Creek Nu c lear Generating Stations/USA John Egdorf Dominion Generation
/Kewaunee Power Station Jack Lewis E ntergy Nuclear/Corporate C. Kelly Walker Operations Support SePrices , Inc. ReYiew Team Chris Boone Southern l'Juclear/Corporate John Callahan Xcel Energy/Corporate/USA Bill Chausse Enereon Services , Inc. Kent Crocker Progress Energy/Brunswick l'Juclear Plant Don Crowl Duke E nergy/Corporate Roger Freeman Constellation Energy l'~uclear Group/Corporate Walt Lee TVA }Juclear/Corporate Ken Meade FE}JOC/Corporate Don Mathena Ne>(tEra Energy/Corporate DaYid Stobaugh EP Consulting , LLC Nick Turner Callaway Plant/STARS Maureen Zav,alick Diablo Canyon Po*,,,*er Plant/STARS NOTICE Neither NEI , nor any of its employees , members , supporting organizations , contractors , or consultants make any 1+1rarranty , eJ.pressed or implied , or assume any legal responsibility for the accuracy or completeness of, or assume any liability for damages resulting from any use of , any information apparatus , methods , or process disclosed in this report or that such may not infringe privately owned rights. M:tcl ea r E n e r t:)" ln s li t1;1 t e, 1 776 1 8-t r ee t N. W , Su it e U H), 14'8s hin g t M D. C. (2()2. 7 3 9.8 ()(}{))
EXECUTl'/E
 
==SUMMARY==
l>le l 99 0 I (RevisieA
: 6) t>ISY6ffi06F 2012 Federal regulations require that a nuclear power plant operator develop a scheme for the classification of emergenC)'
events and conditions. This scheme is a fundamental component of an emergency plan in that it proYides the defined thresholds that 1 , 1 ,ill allow site personnel to rapidly implement a range of pre planned emergency response measures.
An emergency classification scheme also facilitates timely decision making by an Offsite Response Organization (ORO) concerning the imple1:nentation of precautionary or protective actions for the public. The purpose of Nuclear Energy Institute (NEI) 99 01 is to provide guidance to nuclear power plant operators for the development of a site specific emergency classification scheme. The methodology described in this document is consistent vrith Federal regulations , and related US Nuclear Regulatory Commission (NRG) requirements and guidance. In particular , this methodology has been endorsed by the NRG as an acceptable approach to meeting the requirements of 10 CFR § 50.47(b)(4), related sections of 10 CFR § 50, AppendiJ{
E , and the associated planning standard evaluation elements ofNUREG 0654/ FEMA REP 1 , Rev. 1 , Crileri,2 fer Prcper,2tien end E*,1,2luetien
&jRedielegie,2/
Emergenq Respense Pl,21q s cmd Preparedness in Sttpf)ert ef}h1.ele,2r Pe,1 1 er Plents , l>foyember 1980. ~ffiI 99 01 contains a set of generic Initiating Conditions
(!Cs), Emergency Action Levels (EALs) and fission product barrier status thresholds.
It also includes supporting technical basi s information , de*,eloper note s and recommended classification instructions for users. Users should implement
!Cs , EALs and thre s holds that are as close as possible to the generic material presented in this document v ,r ith allov,rance for changes necessary to address site specific considerations such as plant design , location, terminology, etc. Properly implemented , the guidance in *NEI 99 01 v ,r ill yield a site specific emergency classification scheme with clearly defined and readily observable EALs and thresholds. Other benefits include the development ofa sound basis document , the adoption of industry standard instructions for emergenC)' classification (e.g., transient events , classification of multiple events , upgrading , downgrading , etc.), and incorporation of features to improYe human performance. An emergency classification using this scheme will be appropriate to the risk posed to plant *.vorkers and the public , and should be the same as that made by another NEI 99 01 user plant in response to a similar event. The individuals responsible for developing an emergency classification scheme are strongly encouraged to reviev, all applicable NRG requirements and guidance prior to beginning their efforts. Questions concerning this document may be directed to the ~~EI Emergency Preparedness staff , NEI EAL task force members or submitted to the Emergency Preparedness Frequently Asked Questions process. Finally , unique State and local requirements associated with an emergency classification scheme are not reflected in this guidance.
Incorporation of these requirements ma)' be performed on a case by case basis in conjunction with the appropriate ORO agency. Any such changes will require a reviev,r under the applicable sections of IO CFR 50. 
}JE I 99 0 1 (R e~*i sieR 6) }l m*effil:ler 20 1 2 [THIS Pf..:GE IS LEFT BLANK INTENTIONALLY]
II TABLE OF CONTENTS Ne l 99 0 I (R ev i s i oA 6) 1'Jo v eml3 er 2 0 12 EXECUTl'\'li SUM MARY,,, .. ,,,,,, .... ,, .. ,, .. ,,,, .. ,,,, .. ,, .. ,,,, .. ,,,,,,,, .. ,, .. ,,,,,,,,,, .. ,, .. ,,,, .. ,,,, .. ,,,,,, .. ,,,, .. ,,,,,, .... ,,,,,, i 1 REGULATORY BACKGROUND
.................................................................................................
1 1.1 OPERATING REACTORS .........................................................................................................
1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) ......................................................
2 1.3 NRC ORDER EA-12-051
......................................................................................................
4 2 KEY TERMINOLOGY USED IN NEI 99-01 ................................................................................
6 2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) .............................................................................
6 2.2 INITIATING CONDITION (IC) ...................................................................................................
8 2.3 EMERGENCY ACTION LEVEL (EAL) .........................................................................................
8 2.4 FISSION PRODUCT BARRIER THRESHOLD
.................................................................................
8 3 DESIGN OF THE PBNP EMERGENCY CLASSIFICATION SCHEME ........................................
11 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLs) .................................................
11 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS .........................................
17 3.3 PBNP-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION
.........................
18 3.4 IC AND EAL MODE APPLICABILITY
.......................................................................................
20 4 PBNP SCHEME DEVELOPMENT
............................................................................................
23 4.1 GENERAL DEVELOPMENT PROCESS .......................................................................................
23 4.2 CRITICAL CHARACTERISTICS
................................................................................................
24 4.3 INSTRUMENTATION USED FOR EALs .....................................................................................
26 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA
....................................
28 5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS
...................................................
30 5.1 GENERAL CONSIDERATIONS
.................................................................................................
30 5.2 CLASSIFICATION METHODOLOGY
..........................................................................................
32 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS
.........................................................
32 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION
..................................................
32 5.5 CLASSIFICATION OF IMMINENT CONDITIONS
...........................................................................
33 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING
........................................
33 5. 7 CLASSIFICATION OF SHORT-LIVED EVENTS .............................................................................
34 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS
..........................................................................
34 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION
.....................................
35 5.10 RETRACTION OF AN EMERGENCY DECLARATION
......................................................................
35 6 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS .......................................
36 7 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS ..................................
69 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ........................
106 9 FISSION PRODUCT BARRIER ICS/EALS ..............................................................................
109 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ...................
160 11 SYSTEM MALFUNCTION ICS/EALS .....................................................................................
193 APPENDIX A -ACRONYMS AND ABBREVIATIONS
.....................................................................
A-1 Ill 
}lei 99 0 1 (R e~'isieA 6) }Jeveffiser
?Q 12 APPENDIX B -DEFINITIONS
..............................................................
..........................................
B-1 IV l>IE I 99 0 1 (Re~*isieH
: 6) l>le>,effiser 20 1 2 DEVELOMENT OF POINT BEACH EMERGENCY ACTION LEVELS FOR NON PASSl'IE REACTORSBASIS DOCUMENT 1 REGULATORY BACKGROUND
 
===1.1 OPERATING===
 
REACTORS Title 10 , Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities.
Several of these regulations govern various aspects of an emergency cla ssifica tion scheme. A review of the relevant sect ion s listed below will aid the reader in understanding the key termino l ogy provided in Section 3.0 of this document.
* 10 CFR § 50.47(a)(l)(i)
* 10 CFR § 50.47(b)(4)
* 10 CFR § 50.54(q)
* 10 CFR § 50.72(a)
* 10 CFR § 50, Appendix E, IV.B , Assessment Actions
* 10 CFR § 50, Appendix E, IV.C, Activation of Emergency Organization The above regulations are supplemente d by various regulatory guida n ce documents.
Three documents of particular relevance to NEI 99-01 are:
* NUREG-0654/FEMA-REP-l , Criteria/or Pr e paration and Evaluation of Radiological Emergency R espo ns e Plan s and Pr epa redness in Support of N uclear Power Plants, October 1980. [Refer to Appendix 1 , Emergency Action Level Guidelines for Nuclear Pow er Plants]
* NUREG-1022, Event Reporting Guidelines IO CFR § 50. 72 and§ 50. 7 3
* Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors 1 1.2 ~le i 99 0 I (Re\*i s i e A 6) Nevember ?Q 12 The abo 1 t1e list is Aot all iAclusive aAd it is stroAgly recommeAded that scheme developers coAsult *with liceAsiAg/regulatory compliaAee personnel to ideAtify aAd uAderstand all applicable requiremeAts aAd guidaAce.
QuestioAs may also be directed to the NEI EmergeAcy PreparedAess staff. PER:."iiANENTLY DEFUELED STATION NEI 99 01 provides guidaAce for aA emer g eAc y cla ss ificatioA scheme applicable to a permaAeAtly defueled statioA. This is a statioA that geAerated speAt fuel uAder a 10 CFR § 50 liceAse , has permaAeAtly ceased operatioAs aAd will store the speAt fuel oAsite for aA e J fteAded period of time. The emergeAcy cla ss ificatioA levels applicable to this type of statioA are coAsisteAt with the requiremeAts of 10 CFR § 50 aAd the guidaAce iA NUREG 0654/FEMA REP 1. In order to rela,c the emergency plaA requ i remeAts applicable to aA operating statioA , the owner of a permaneAtly defueled statioA mu s t demoAstrate that no credible event caA result in a sigAificaAt radiological relea s e be)'OAd the site bouAdary.
It is e>fpeeted that this verificatioA will coAfirm that the source term aAd motive force available iA the permaAeAtly defueled coAditioA are iAsufficieAt to warraAt classifications of a Site Area EmergeAC)' or GeAeral En~ergeAcy.
Therefore , the geAeric IAitiatiAg CoAditioAs (ICs) aAd EmergeAcy Action Levels (EALs) applicable to a permaAeAtly defueled statioA may result in either a Notification of Unusual EveAt (NOUE) or aA Alert c l assificat i oA. The generic !Cs aAd EALs are preseAted in AppeAdi>c C , Pcrmcmc11Hy Defitckd Stc1tie11 ICs/EALs. H.12._INDEPENDENT SPENT FuEL STORAGE INSTALLATION (ISFSI) Selected guidance in NEI 99-01 is applicable to licensees electing to use their 10 CFR 50 emergency plan to fulfill the requirement s of l O CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR f-50 and the guidance in NUREG 0654/FEMA-REP-1. The initiating conditions germane to a 10 CFR-§ 72.32 emergency plan (as described in NUREG-1567) are subsumed within the classification scheme for a 10 CFR-§--50.47 emergency plan. T he generic ICs and EALs for an ISFSI are presented in Section 8 , ISFSI ICs/EALs. IC E-HU l covers the spectrum ofcredible natural and man-made events included within the scope of an ISFSI de sign. This IC is not applicable to installations or facilities that may process and/or repackage spent fuel (e.g., a Monitored Retrievable Storage Facility or an ISFSI at a spent fuel processing facility).
In addition, appropriate aspects ofIC HUI and IC HAI should also be included to address a HOSTILE ACTION directed against an ISFSI. The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, A R egu latory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactiv e Material Licensees.
NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has 2 l>Jel 99 0 I (RevisieA
: 6) l>Je~*emser 2012 insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effec tive Dose Eq uivalent.
Regarding the above information , the expectations for an offsite response t o an Alert classified under a 10 CFR-§-72.32 emergency plan are generally consistent with those for aan Notification of Unusual EventUnusual Event in a 10 CFR-§-50.47 emergency plan (e.g., to provide assistance if requested).
Also , the licensee's Emerge nc y Response Organization (ERO) required for 10 CFR -§-72.32 emergency plan is different than that pr esc ribed for a 10 CFR-§-50.47 emergency plan ( e.g., no emergency technica l support function).
3 
+-All_NRC ORDEREA-12-051
~lei 99 0 I (RevisieA
: 6) ~Jeveffiser 2012 The Fukushima Daiichi accident of March 11, 2012, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of co re cooling and containment integrity , and ultimately led to core damage in three reactor s. While the loss of power also impaired the spent fuel pool cooling function , sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling. Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule , 10 CPR 50.109(a)(4)(ii).
Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliabl e Spent Fuel Pool Instrumentation, on March 12 , 2012, to all US nuclear plants with an operating license , construction permit , or combined construction and operating license. NRC Order EA-12-051 states, in part, "All licensees
... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel:
(1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck , and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." To this end, all licensees must provide:
* A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;
* A display in an area accessible following a severe event; and
* Independent electrical power to each instrument channel and provide an alternate remote power connection capability. NEI 12-02, Industry Guidance for Compliance with N R C Order EA-12-051, " To Modify Licenses with Regard to Reliabl e Spent Fuel Pool In strume ntation ," provides guidance for complying with NRC Order EA-12-051. NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051.
These EALs are included within e>C.isting IC~ ~RA2 , and new ICs A82 RS2. and AmRG2. Associated EAL notes , bases and developer notes are also provided.
It is recommended that these EALs be implemented when the enhanced spent fuel pool level instrumentation is available for use. 4 1.5 }JE I 9 9 0 I (ReYisieA 6) }l evember 2012 !he repilator y process that licensees follov,r to ma , mcludmg non scheme changes to EAL . 10 ke changes to their emergency plan I * . n S , IS CFR 50 54E ) I d ' regu at1on , licensees are responsible f; " 1 .* q .n accor ance with this .. 'heth . ~r e, a uatmg a proposed ch d d v_r ef" 1 or not 1t results in a reduction in the effe .* , ange anetef"mining licensees determination the licensee "'ill " th ct~reness of the plan. As a result of the :{:; * ' ,r e1 erma'ethe h or pnor review and appro''al in acco d . h ' c ange or submit it to the l>lRC r r ance wit IO CFR 50.90. l'..,PPLICABILITY TO ADVANCED AND S MALL MODULAR REACTOR DESIGNS The gu_idance in this document primarily addresse . the Un1~ed States , operating or permanently de& ; :omm:rc1al nuclear po*.ver reactors in generation plant designs)* home"e *t ue e , as o 2012 Eso called 1 5*--ane--2-ftti E ft ' ,r
* r , 1 may be adapt d t d o en referred to as 3m generation plant d ~eo a vanced non passive designs cla s sification scheme for an ad"anced es1gn~) as well. Developers of an emergency deviations from the generic gui~an t non passive reacto~ plant may need to propose d . . ce o account for the dtffi
* d
* an cntena , and operating characteristics and b T~erences mes1gn;farameters .nl..a..n.t.c_
capa 1 1t1es between 2ftti ~-,.1 ,,
* pm= ' generation There are significant design and operating differe . pow I E f nces betuceen la : ,ref p antso any generation) and Small M d ,"rge commercial nuclear m source term). For this reason this do
* o . u ar Reac:ors ESMRs) Ee.g., differences
' cument is not applicable to SHR . s. 5 2 KEY TERMINOLOGY USED IN NEI 99-01 NEI 99 0 I (RevisieA
: 6) }levember 2012 There are several key terms that appear throughout the NEI 99 OIEAL methodology. These terms are introduced in this section to support understanding of subsequent material.
As an aid to the reader , the following table is provided as an overview to illustrate the relationship of the terms to each other. Unusual Event Initiating Condition Emergency Action Level (1)
* Operating Mode Applicability
* Notes
* Basis I Emergency Classification Level Alert I SAE Initiating Condition Emergency Action Level (1)
* Operating Mode Applicability
* Notes
* Basis Initiating Condition E mergency Action Level (1)
* Operating Mode Applicability
* Notes
* Basis I GE Initiating Condition Emergency Action Level (1)
* Operating Mode Applicability
* Notes
* Basis (1) -When making an emergency classification , the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition.
This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information.
In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EA L:-~ 2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or cond ition s accord in g to (1) potential or actual effects or consequences , and (2) resulting onsite and offsite response actions. The emergency classification levels , in ascending order of severity, are: * *Notification of Unusual EventUnusual Event (N OUEUE)
* Alert
* Site Area Emergency (SAE)
* General Emergency (GE) I 2.1.l Notification of Unusual EventUnusual Event (NOUEUE)+ Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systemsSAFETY SYSTEMS occurs. + This term is semetimes shertened te UAusual E¥eAt (UE) er ether similar site speeifie terminelegy.
The terms }letifieatien sf Unusual e>,*ent , NOUe aAd UAusual £,*ent are used interehangeably th.eugheut this deeument 6 l'lf:I 99 0 I (ReYisien
: 6) l'leveffiser 2012 Purpose: The purpose of this classification is to assure that the first step in future response has been carried out , to bring the operations staff to a state of readiness , and to provide systematic handling of unu s ual event information and decision-making.
 
====2.1.2 Alert====
E vents are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the E PA PAG exposure levels. Purpose: The purpose of this classification is to assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring if required , and provide offsite authorities current information on plant status and parameters.
2.1.3 Site Area E mergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to , equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.
Purpose: The purpose of the Site Area Emergency declaration is to assure that emergency response centers are staffed, to assure that monitoring teams are dispatched , to assure that personnel required for evacuation of near-site areas are at duty stations if the situation becomes more serious , to provide consultation with offsite authorities , and to provide updates to the public through government authorities.
 
====2.1.4 General====
Emergency (GE) Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Purpose: The purpose of the General Emergency declaration is to initiate predetermined protective actions for the public , to provide continuous assessment of information from the licensee and offsite organizational measurements , to initiate additional measures as indicated by actual or potential releases , to provide consultation with offsite authorities , and to provide updates for the public through government authorities.
7 
' 2.2 INITIATING CONDITION (IC) }lel 99 0 1 (Re~'isieA
: 6) }le Yeffieer 2012 An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.
Discussion:
An IC describes an event or condition , the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed as a continuous, measurable parameter (e.g., RCS leakage), an event (e.g., an earthquake) or the status of one or more fission product barriers ( e.g., loss of the RCS barrier).
Appendix 1 of NUREG-0654 does not contain example Emergency Action Levels_ (EALs) for each ECL , but rather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency , or events that could lead to a radiological emergency , has occurred).
NUREG-0654 states that the Initiating Conditions form the basis for establishment by a licensee of the specific plant instrumentation readings (as applicable) which , if exceeded , would initiate the emergency classification.
Thus , it is the specific instrument readings that would be the EALs. Considerations for the assignment of a particular Initiating Condition to an emergency classification level are discussed in Section 3. 2.3 EMERGENCY ACTION LEVEL (EAL) A pre-determined , site-specific , observable threshold for an Initiating Condition that , when met or exceeded , places the plant in a given emergency classification level. Discussion:
EAL statements may utilize a variety of criteria including instrument readings and status indications
; observable events; results of calculations and analyses; entry into particular procedures
; and the occurrence of natural phenomena.
 
===2.4 FISSION===
PRODUCT BARRIER THRESHOLD A pre-determined, site-specific , observable threshold indicating the loss or potential loss of a fission product barrier. Discussion:
Fission product barrier thresholds represent threat s to the defense in depth design concept that precludes the release of radioactive fission products to the environment.
This concept 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 barrier s are:
* Fuel Clad
* Reactor Coolant System (RCS)
* Containment
---Upon determination that one or more fission product barrier thresholds have been exceeded, the combination of barrier loss and/or potential loss thresholds is compared to the fission product barrier IC/EAL criteria to determine the appropriate ECL. In some accident sequences , the !Cs and E ALs presented in the Abnormal Radiation Levels/ Radiological Effluent (AR) Recognition Category will be exceeded at the same 8 NBI 99 01 (RevisieA
: 6) J>JeYemller 2012 time , or shortly after , the loss of one or more fission product barriers.
This redundancy is intentional as the former ICs address radioactivity releases that result in certain offsite doses from whatever cause, including events that might not be fully encompassed by fission product barriers ( e.g., spent fuel pool accidents , design containment leakage following a LOCA, etc.). 9 10 M;: 1 99 01 (ReYisieA
: 6) J>levemaer 2012 
}le i 99 0 I (Re\*i s i o A 6) }lovelf!eer 2012 3 DESIGN OF THE NEI 98 01PBNP EMERGENCY CLASSIFICATION SCHEME 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) An effective emergency classification scheme must incorporate a realistic and accurate assessment of risk , both to plant workers and the public. There are obvious health and safety risks in underestimating the potential or actual threat from an event or condition; however, there are also risks in overestimating the threat as well (e.g., harm that may occur during an evacuation).
The NEI 99 OlPBNP emergency classification scheme attempts to strike an appropriate balance between reasonably anticipated event or condition consequences , potential accident trajectories, and risk avoidance or minimization.
There are a range of " non-emergency events" reported to the US Nuclear Regula to ry Commission (NRC) staff in accordance with the requirements of 10 CFR -§-50. 72. Guidance concerning these reporting requirements , and example events, are provided in NUREG-1022.
Certain events reportable under the provisions of 10 CFR -§-50. 72 may also require the declaration of an emergency.
In order to align each Initiating Conditions (IC) with the appropriate ECL , it was necessary to determine the attributes of each ECL. The goal of this process is to a nswer the question, "What events or conditions should be placed under each ECL ?" The following sources provided information and context for the development of ECL attributes.
* Assessments of the effects and consequences of different types of events and conditions
* Typical PBNP abnormal and emergency operating procedure setpoints and transition criteria
* Typical PBNP Technical Specification limits and controls
* Radiological EftlueAt Teclrnical 8pecificatioAs (RET8)/0ffsite Dose Calculation Manual (ODCM) radiological release limits
* Review of selected Updated Final Safety Analysis Report (UFSAR) accident analyses
* Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs)
* NUREG 0654, Appendix 1 , Emergency Action Level Guidelines for Nuclear Pow er Plants
* Industry Operating Experience
* Input from iAdustry PBNP subject matter experts aAd NRG staff members The following ECL attributes were created by the RevisioA 6 PreparatioA Team to aid in the development of I Cs and Emergency Action Levels (EALs ). The team decided to iAclude the attributes iA this revisioA siAce theyThe attributes may be useful in briefing and training settings ( e.g., helping an Emergency Director understand why a part i cular condition is classified as an Alert). It should be stressed that developers Rot attempt to redefiAe these attributes or apply them iA aAy fashioA that would chaAge the geAeric .d . d. h. d + gu1 aAce ooAtame m t 1socumeAt
., + The 1:1 s e of eCL attri01:1tes is at the EliseretioA of a I ieeAsee aAEI i s Aot a req1:1irelf!eAt of the NRG. If a lieeAsee ehooses iA iAeorporate the eCL attri01:1tes iAto their s e heA~e ea s i s El o e1:1lf!eAt , it lf!1:1 s t ee very elear that the }IRC staff llas R o t eAElorseEI their aeeeptaeilit y or applieati o A for aA y p1:1rp os e. TA parti61:llar , the staffEloes Rot eoAsiEler the 11 
}JEI 99 01 (Re\*isieA 6) }levemeer 2012 a~riettte statemeAts te st119erseEle tke estaeliskeEI EGL ElefiAitieAs.
As a resttlt , tke ttse ef tke attrietttes as a easis fer *ttstif:,*iag EAL ckaages is t1Aacce19taele. 12 The attributes of each BCL are presented below. 13 }JEI 99 01 (RevisieA
: 6) }JeveA'leer 2012 
 
====3.1.1 Notification====
 
of Unusual EventUnusual Event (N OUEUE) J>JE;I 99 0 I (Re\*isieR 6) J>le\*ember 2012 A n Notification of Unusual EventUnusual Event , as defined in section 2.1.1 , includes but is not limited to an event or condition that involves: (A)A precursor to a more significant event or condition. (B) A minor loss of control of radioactive materials or the ability to control radiation levels within the plant. (C) A consequence otherwise significant enough to warrant notification to local, State and Federal authorities.
 
====3.1.2 Alert====
An Alert , as defined in section 2.1.2 , includes but is not limited to an event or condition that involves: (A)A loss or potential loss of either the fuel clad or Reactor Coolant System (RCS) fission product barrier. (B) An event or condition that significantly reduces the margin to a loss or potential loss of the fuel clad or RCS fission product barrier. (C) A significant loss of control of radioactive materials resulting in an inability to control radiation levels within the plant , or a release of radioactive materials to the environment that could result in doses greater than 1 % of an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the OWNER CONTROLLED AREA , including those directed at an Independent Spent Fuel Storage Installation (ISFSI). I 3 .1.3 Site Area Emergency (SAE) A Site Area Emergency , as defined in section 2.1.3, includes but is not limited to an event or condition that involves: (A) A loss or potential loss of any two fission product barriers -fuel clad , RCS and/or containment. (B) A precursor event or condition that may lead to the loss or potential loss of _multiple fission product barriers within a relatively short period of time. Precursor events and conditions of this type include those that challenge the monitoring and/or control of multiple safety systemsSAFETY SYSTEMS. (C) A release of radioactive materials to the environment that could result in doses greater than 10% of an EPA PAG at or beyond the site boundary. (D) A HOSTILE ACTION occurring w ithi n the plant PROTECTED AREA. 14 3 .1.4 General Emergency (GE) }IE;I 99 01 (RevisieA
: 6) }levember 2012 A General Emergency, as defined in section 2.1.4, includes but is not limited to an event or condition that involves: (A) Loss of any two fission product barriers AND loss or potential loss of the third barrier -fuel clad, RCS and/or containment. (B) A precursor event or condition that, unmitigated, may lead to a loss of all three fission product barriers.
Precursor events and conditions of this type include those that lead directly to core damage and loss of containment integrity. (C) A release of radioactive materials to the env ironm ent that could result in doses greater than an E PA PAG at or beyond the site boundary. (D)A HOSTILE ACTION resulting in the Joss of key safety functions (reactivity control, core cooling/RPV water level or RCS heat removal) or damage to spent fuel. 15 3 .1.5 Risk-Informed Insights }/el 99 0 I (RevisieR
: 6) }/eY0ffi80f 20 J 2 Emergency preparedness is a defense-in-depth measure that is independent of the assessed risk from any particular accident sequence; however , the development of an effective emergency classification scheme can benefit from a review of risk-based assessment results. To that end , the development and assignment of certain ICs and EALs also considered insights from several site-specific probabilistic safety assessments (PSA also knovm as probabilistic risk assessment, PRA). Some generic insights from this review included:
: 1. Accident sequences involving a prolonged loss of all AC power are significant contributors to core damage frequency at many Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). For this reason , a loss of all AC power for greater than 15 minutes, with the plant at or above Hot Shutdown, was assigned an ECL of Site Area Emergency.
Precursor events to a loss of all AC power were also included as an Unusual Event and an Alert. A station blackout coping analyses performed in response to 10 CFR -§-50.63 and Regulatory Guide 1.155 , Station Blackout , may be used to determine a time-based criterion to demarcate between a Site Area Emergency and a General Emergency.
The time dimension is critical to a properly anticipatory emergency declaration since the goal is to maximize the time available for State and local officials to develop and implement offsite protective actions. 2. For severe core damage events, uncertainties exist in phenomena important to accident progressions leading to containment failure. _Because of these uncertainties , predicting the status of containment integrity may be difficult under severe accident conditions.
_This is why maintaining containment integrity alone following sequences leading to severe core damage is an insufficient basis for not escalating to a General Emergency.
: 3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containment bypass, a large Loss of Coolant Accident (LOCA) with early containment failure, a Station Blackout lasting longer than the site specificPBNP coping period , and a reactor coolant pump seal failure. The generic EAL methodology needs to be sufficiently rigorous to address these sequences in a timely fashion. 16 l>lel 99 O 1 (RevisieA
: 6) l>leveFAber
?Q 12 3 .2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS The NEI 99-0 I methodology makes use of symptom-based, barrier-based and based ICs and EALs. E ach type is discussed below. Symptom-based ICs and EALs are parameters or conditions that are mea s urable over some range using plant instrumentation (e.g., core temperature , reactor coolant level , radiologic a l effluent , etc.). When one or more of these parameters or conditions are normal , reactor operators will implement procedures to identify the probable cause(s) and take corrective action. Fission product barrier-based ICs and EALs are the subset of symptom-based EALs that refer specifically to the level of challenge to the principal barriers against the release of radioactive material from the reactor core to the environment.
These barriers are the fuel cladding , the reactor coolant system pressure boundary , and the containment.
The barrier=-based I Cs and EALs consider the level of challenge to each individual b a rrier -potentially lost and lost -and the total number of barriers under challenge.
Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance.
These include the failure of an automatic reactor scram/trip to shut down the reactor , natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release. 17 3.3 NSSS DESIGN DIFFERENCES J>J E I 99 0 I (RevisieA
: 6) November 2012 The l'ffil 99 01 emergency classificat i on scheme accounts for the design differences bet\Yeen PWRs and B\1/Rs by specifying EALs unique to each type of Nuclear Steam Supply System (NSSS). There are also significant design differences among PWR NSSSs; therefore , guidance is provided to a i d in the development of EALs appropriate to different PWR NSSS types. Where necessary , development guidance also addresses unique considerations for advanced non passive reactor designs such as the Advanced Boiling Water Reactor (ABWR), the Advanced Pressurized Water Reactor (/\PWR) and the Evolutionary Povrer Reactor (BPR). Developers ,viii need to consider the re l evant aspects of their plant's design and operating characteristics when converting the generic guidance of this document into a site specific classification scheme. The goal is to maintain as much fidelity as possible to the intent of generic ICs and EALs with i n the constraints imposed by the plant design and operating characteristics.
To this end , developers of a scheme for an advanced non passive reactor may need to add , modify or delete some information contained in this document; these changes will be reviev,'ed for acceptability by the NRG as part of the scheme appro,,al process. The guidance in l'ffil 99 01 is not applicable to advanced passive light 'Nater reactor designs. AA Emergency Classification Scheme for this type of p l ant shou l d be developed in accordance with l'ffil 07 01 , Methetielegy fer Develepment
&}Emergency*
Aetien Levels , Ad1>1 tmeed Pt1ssi 1 re Light W~tcr Ret1eters. 4 3 .3 PBNP-SPECI FIC 0RGANIZA TION AND PRESENTATION OF GENERIC INFORMATION The scheme's generic information is organized by Recognition Category in the following order.
* A-R.-Abnormal Radiation Levels/ Radiological Effluent-Section 6
* C -Cold Shutdown / Refueling System Malfunction
-Section 7
* E -Independent Spent Fuel Storage Installation (ISFSI) -Section 8
* F -Fission Product Barrier -Section 9
* H -Hazards and Other Conditions Affecting Plant Safety -Section 10
* S -System Malfunction
-Section 11
* PD Permanently Defueled Station AppendiJ, C E ach Recognition Category section contains a matrix showing the ICs and their associated emergency classification levels. The following information and guidance is provided for each IC:
* ECL -the assigned emergency classification level for the IC.
* Initiating Condition
-provides a summary description of the emergency event or condition.
* Operating Mode Applicability
-Li s ts the modes during which the IC and associated EAL(s) are applicable (i.e., are to be used to classify events or conditions).
18 l>lf.l 99 0 I (RevisieA
: 6) l>levember 2012
* Examf)le Emergency Action Leve l (s)-Provides examples of reports and indications that are considered to meet the intent of the IC. Developers should address each example EAL. If the generic approach to the development of an eM,ample EAL cannot be used (e.g., an assumed instrumentation range is not available at the plant), the developer should attempt to specify an alternate means fur identifying entry into the IC. For Recognition Category F , the fission product barrier thresholds are presented in tables applicable to BWRs and P\VRs , and arranged by fission product barrie r and the degree of barrier challenge (i.e.,_-potential loss or loss). This presentation method shows the synergism among the thresholds , and supports accurate assessments.
* Basis -Provides background information that explains the intent and application of the IC and EALs. In some cases, the basis a l so includes relevant source information and references. 19 NEI 99 O I (Re~*isioR 6) }loYeFFther
?Q 12
* De¥eleper Netes Information that supports the development of the site specific ICs and EALs. This may include clarifications, references , examples , instructions for calculations , etc. Developer notes should not be included in the site's emergency classification scheme basis document.
Developers may elect to include information resulting from a developer note action in a basis section.
* ECL Assigemeet Attributes Located within the Developer Notes section , specifies the attribute used for assigning the IC to a given EGL. H3.4 IC AND EAL MODE APPLICABILITY The "t>IBI 99 01 PBNP emergency classification scheme was developed recogn i zing that the applicability ofICs and EALs will vary wit h p l ant mode. For example, some symptom-based ICs and EALs can be assessed on l y during the power operations, startup , or hot standby/s h utdown modes of operation when all fission product barriers are in place, and plant instrumentation and safety systemsSAFETY SYSTEMS are fully operational.
In the cold shutdown and refueling modes, different symptom-based ICs and EALs will come into play to reflect t h e opening of systems for routine maintenance , the unavailability of some safety systemSAFETY SYSTEM components and the use of alternate instrumentation.
The following tab l e shows which Recognition Categories are applicable in each p l ant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes. MODE APPLICABILITY MATRIX Recog ni tion Category Mode AR C E F H s Power Operations X X X X X Startup X X X X X Hot Standby X X X X X Hot Shutdown X X X X X Co l d Shutdown X X X X Refueling X X X X Defueled X X X X Permanently
' Defueled , 20 MODE 1 2 3 4 5 6 NIA Tvaieal BWR OaeFRting Mades Pov,cer 013erations (1): Mode Sv,*itch in Run NEIi 99 01 (ReYisieR
: 6) J>Jevemeer 2012 Startu13 (2): Mode Switch in Startu13/Hot Standby or Refuel (with all vessel head bolts fully tensioned)
Hot Shutdown (3): Mode S*witch in Shutdown, Average Reactor Coolant Temperature
>200 °F Cold Shutdovm (4): Mode s,.,,*itch in Shutdown, Average Reactor CoolaHt Tem13erature 200 °P Refueling (5): Mode Switch in Shutdov,*n or Refuel, aHd one or more vessel head bolts less than fully tensioned.
Typieal PWR PBNP OPERATING MODES TITLE REACTIVITY
% RATED AVERAGE CONDITION THERMAL REACTOR .cKilll POWER(a) COOLANT TEMPERATURE rEl Power Ooeration
>0.99 >5 NA Startuo >0.99 <5 NA Hot Standbv <0.99 NA >350 Hot Shutdown<0> <0.99 NA 350 > T avn > 200 Cold Shutdown <0> <0.99 NA <20 0 Refueling le> NA NA NA Defueled All fuel removed from the reactor vessel (full core offload during refueling or extended outage) (a) Excluding decay heat (b) All reactor vessel head closure bolts fully tensioned. ( c) One or more reactor vessel head closure bolts les s than fullv tensioned.
Power Operations (1): Reactor Pov,*er > 5%, Keff > 0.99 Startu13 (2): Hot Standby (3): Reactor Power< 5%, Keff~ 0.99 RCS~ 350 &deg;P, Keff< 0.99 Hot Shutdown (4): 200 &deg;P <RCS< 350 &deg;P, Keff < 0.99 Cold Shutdovm (5): RCS< 200 &deg;P, Keff < 0.99 Refueling (6): One or more vessel head closure bolts less than fully tensioned Developers will need to incor13orate the mode criteria from unit specific Technical 21 l>lEI 99 01 (Re\'isieA 6) l>ie&#xa5;ember 2012 Specifications into their emergency classification scheme. In addition, the scheme must also include the follov,ing mode designation specific to NEI 99 01: Defueled (}>fone):
All fuel removed from the reactor vessel (i.e., full core offload during refueling or e}(tended outage). 22 
}Jel 99 01 (Re&#xa5;isieB
: 6) }le~*emeer 2012 4 SITE &PECIFICPBNP SCHEME DEVELOPMENT GUIDANCE This section provides detailed guidance for developing a site specific emergency classification scheme. Conceptually, the approach discussed here mirrors the approach used to prepare emergency operating procedures generic material prepared by reactor *,ender ovmers groups is converted by each nuclear po 1 Ner plant into site specific emergency operating procedures. Like\vise, the emergency classification scheme developer
*.viii use the generic guidance in NEJ 99 01 to prepare a site specific emergency classification scheme and the associated basis document.
It is important that the NEI 99 01 emergency classification scheme be implemented as an integrated package. Selected use of portions of this guidance is strongly discouraged as it 1 Nill lead to an inconsistent or incomplete emergency classification scheme that 1 ,vill likel)' not receiYe the necessary regulatory approval.
 
===4.1 GENERAL===
IMPLEMENTATION CUIDAJ!WEDEVELOPMENT PROCESS The guidance in NEI 99 01 is not intended to be applied to plants "as is"; however, deYelopers should attempt to keep their site specific schemes as close to the generic guidance as possible.
The goal is to meet the intent of the generic Initiating Conditions (ICs) and Emergency Action LeYels (EALs) 1tvithin the context of site specific characteristics locale , plant design, operating features , terminology, etc. Meeting this goal *will result in a shorter and less cumbersome NRG reYiew and approYal process , closer alignment with the schemes of other nuclear pov,er plant sites and better positioning to adopt future industr)' v>'ide scheme enhancements. *when properly deYeloped , the The PBNP ICs and EALs should were developed to be unambiguous and readily assessable.
As discussed in Section 3 , the generic guidance includes ICs and e~rnmple EALs. It is the intent of this guidance that .!2Q!h be included in site specific documents as each serYes a specific purpose. The IC is the fundamental event or condition requiring a declaration.
_The EAL(s) is the pre-determined threshold that defines when the IC is met. If some feature of the plant location or design is not compatible with a generic IC or EAL, efforts should be made to identify an alternate IC or EAL. If an IC or EAL includes an explicit reference to a mode dependent technical specification limit that is not applicab l e to the plant, then that IC and/or EAL need not be included in the site specific scheme. In these cases, deYelopers must proYide adequate documentation to justify *.vhy the IC and/or EAL *.vere not incorporated (i.e., sufficient detai l to allow a third party to understand the decision not to incorporate the generic guidance). Useful acronyms and abbreviations associated with the NEI 99 01 PBNP emergency classification scheme are presented in Appendix A, Acronyms and Abbreviations. specific entries may be added if necessary. 23 
}/E l 99 01 (Re~*i s ioA 6) }Joyeffiber JO 12 Many words or terms used in the NEI 99 OlPBNP emergency classification scheme have scheme-specific definitions.
These words and terms are identified by being set in all capital letters (i.e., ALL CAPS). The definitions are presented in Appendix B , Definitions.
Below are examples of acceptable modifications to the generic guidance.
These may be incorporated depending upon site developer and user preferences.
* The ICs *within a Recognition Category may be placed in reverse order for presentation purposes (e.g., start with a General Emergency at the left/top of a user aid , followed by Site Area Emergency , Alert and NOUE).
* The Initiating Condition numbering may be changed.
* The first letter of a Recognition Category designation may be changed , as follows , provided the change is carried through for all of the associated IC identifiers.
* R may be used in lieu of A
* M may be used in lieu of 8 For e1rnmple , the Abnormal Radiation Le*,rels / Radiological Effluent category designator
" A" (for Abnormal) may be changed to " R" (for Radiation).
This means that the associated ICs would be changed to RUl , RU2 , RAl , etc.
* The ICs and EALs from Recognition Categories 8 and C may be incorporated into a common presentation method (e.g., one table) provided that all related notes and mode applicability requirements are maintained.
* The ICs and EALs for Emergency Director judgment and security related events may be placed under separate Recognition Categories.
* The terms EAL and threshold may be used interchangeably.
The material in the Developer Notes section is included to assist developers with crafting correct IC and EAL statements. This material is not required to be in the final emerge0cy classification scheme basis document.
 
===4.2 CRITICAL===
CHARACTERISTICS As discussed above, developers are encouraged to keep their site specific schemes as close to the generic guidance as possible.
When crafting the scheme, developers should satisfy themselvesPBNP ensured that certain critical characteristics have been met. These critical characteristics are listed below.
* The ICs, EALs, Operating Mode Applicabi lit y criteria, Notes and Basis information are consistent with industry guidance; while the actual wording may be different, the classification intent is maintained.
With respect to Recognition Category F, specific scheme must PBNP includes a some type of user-aid to facilitate timely and accurate classification of fission product barrier losses and/or potential losses. The user-aid logic must beis consistent with the classification logic presented in Section 9. 24 
}lei 99 0 I (RevisieR
: 6) }levember 20 1 2
* The ICs , EALs , Operating Mode Applicability criteria , Notes and Basis information are technically complete and accurate (i.e., they contain the information necessary to make a correct classification).
* EAL statements use objective criteria and observable values.
* ICs , EALs , Operating Mode Applicability and Note statements and formatting consider human factors and are user-friendly.
* The scheme facilitates upgrading and downgrading of the emergency classification where necessary.
* The scheme facilitates classification of multiple concurrent events or conditions.
25 1>11;1 99 01 (Revi s i e R 6) T>l eveffiser 2012 4.3 INSTRUMENTATION USED FOR EALS 4.4 lllstrumentation referenced in EAL statements should include that described in the emergency plan section 1.vhich addresses 10 CFR 50.47(b)(8) and (9) and/or Chapter 7 of the FSA.R. Instrumentation used for EALs need not be safet)' related , addressed by a Technical Specification or ODCM/RETS control requirement , nor po,vered from an emergency power source; however , EAL developers should strive to PBNP incorporate g_ instrumentation that is reliable and routinely maintained in accordance with site programs and procedures.
Alarms referenced in EAL statements should beare those that are the most operationally sign ificant for the described event or condition. Scheme developers should ensure that specified values used as EAL setpoints are within the calibrated range of the referenced instrumentation , and consider any automatic instrumentation functions that may impact accurate EAL assessment.
In addition , EAL setpoint values should do not use terms such as "off-sca le low" or " off-scale high" since that type of reading may not be readily differentiated from an instrument failure. Findings and violations related to E,.\L instrumentation issues may be located on the :NRG *.vebsite.
PRESENTATION OF SCHEME INFORMATION TO USERS The US :Nuclear Regulator)
' Commission (1'tRC) e)C:pects licensees to establish and maintain the capability to assess , classify and declare an emergency condition promptly v,ithin 15 minutes after the availability of indications to plant operators that an emergency action level has been , or may be , exceeded.
When writing an emergency classification procedure and creating related user aids , the developer must determine the presentation method(s) that best supports the end users by facilitating accurate and timely emergency classification. To this end, developers should consider the following points.
* The first users of an emergency classification procedure are the operators in the Control Room. During the allowable classification time period , they may have responsibil i ty to perform other critical tasks , and will likely have minimal assistance in making a classification assessment.
* As an emergency situation evolves, members of the Control Room staff are likely to be the first personnel to notice a change in plant conditions.
They can assess the changed conditions and, when 1 ,varranted , recommend a different emergency classification level to the Technical Support Center (TSC) and/or Emergency Operations Facility (EOF).
* Emergency Directors in the TSC and/or EOF will have more opportunity to focus on making an emergency classification , and will probably have advisors from Operations available to help them. Emergency classification scheme information for end users should be presented in a manner with which licensed operators are most comfortable. Developers
*will need to work closely with representatives from the Operations and Operations Training Departments to develop readily usable and easily understood classification tools (e.g., a procedure and related user aids). If necessary , an alternate method for presenting 26 4.5 }Jel 99 01 (Re,,*isieR 6) }Je,,*emeer 2012 emergenc)'
classification scheme information may be developed for use b)' Emergency Directors and/or Offsite Response Organization personnel.
A *Nallboard is an acceptable presentation method provided that it contains all the information necessary to make a correct emergency elassification. This information ineludes the ICs, Operating Mode Applicability criteria , EALs and Notes. Notes may be kept with each applicable EAL or moved to a common area and referenced; a reference to a Note is acceptable as long as the information is adequately captured on the 'Nallboard and pointed to by each applicable EAL+. Basis information need not be included on a wallboard but it should be readily available to emergenC)'
elassification decision makers. In some cases, it may be advantageous to develop two wallboards one for use during pmver operations, startup and hot conditions, and another for cold shutdo*Nn and refueling conditions.
Alternative presentation methods for the Recognition Category F ICs and fission product barrier thresholds are acceptable and inelude flow charts, block diagrams, and checklist tyf)e tables. Developers must ensure that the site specific method addresses all possible threshold combinations and elassification outcomes shown in the BWR or PWR EAL fission product barrier tables. The NRG staff considers the presentation method of the Recognition Category F information to be an important user aid and may request a change to a particular proposed method if , among other reasons, the change is necessary to promote consistency across the industry. INTEGRATION OF ICs/EALs V/ITII PLANT PROCEDURES A rigorous integration of IC and E/\L references into plant operating procedures is not recommended.
This approach would greatly increase the administrative controls and v,rorkload for maintaining procedures.
On the other hand, performance challenges may occur if recognition of meeting an IC or EAL is based solely on the memory of a licensed operator or an Emergency Director, especially during periods of high stress. Developers should consider placing appropriate visual cues (e.g., a step, note, caution, etc.) in plant procedures alerting the reader/user to consult the site emergency classification procedure.
Visual cues could be placed in emergency operating procedures, abnormal operating procedures , alarm response procedures, and normal operat i ng procedures that apply to cold shutdown and refueling modes. As an e~rample, a step, note or caution could be placed at the beginning of an RCS leak abnormal operating procedure that reminds the reader that an emergency classification assessment should be performed.
+ Where Bflflref)riate , the }Jetes shewH iR the geRerie g1:1iElaRee t)'flieally iRelt1Ele the eveHt J eeRElitieH ECL aREI the E11:1ratieA time SfleeifieEI iR the EAL. If EleYelefJers flrefer te ha,,*e se,*eral JCs refereRee a eemmeR }IOTE eR a ,.,,*al l eearEI Elisfllay , it is aeeef)taele te remeYe the EGL aREI time eriterieR aAEI 1:1se a geRerie statemeRt.
Fer e1camflle , a ee1ttmeA NOTE ee1:1IEI reaEI "The EmergeAe)'
Direeter she1:1IEI Eleelare the emergeRey flF0mfltly 1:tfleR EletermiRiRg that the Bflfllieaele EAL time has eeeA e){eeeEleEI , er 1 Nill l ikely ee e1,eeeEleEI
." 27 
 
===4.6 BASIS===
DOCUMENT r>J e l 99 0 I (RevisieA
: 6) r>Je~'effiser ?Q 12 A basis document is an integral part of an emergency classification scheme. Tile material in this document supports proper emergency classification decision making by providing informing background and develo13ment information in a readily accessible format. It can be referred to in training situations and when making an actual emergenc;'
classification, if necessary.
Tile document is also useful for establishing configuration management controls for EP related equi13ment and e>(f)laining an emergency classification to offsite authorities.
Tile content of tile basis document should include , at a minimum, tile following:
* A site s13ecific Mode A1313licability Matri>( and descri13tion of 013erating modes , similar to that 13resented in section 3.5.
* A discussion of tile emergency classification and declaration 13rocess reflecting tile material 13resented in Section 5. This material may be edited as needed to align with site s13ecific emergency 13lan and im13lementing 13rocedure requirements.
* Eaoll Initiating Condition along with tile associated EALs or fission 13roduct barrier thresholds, 013erating Mode A1313licabilit;', J>Jotes and Basis information.
* A listing of acronyms and defined terms , similar to that 13resented in A1313endices A and B , respeetivel;
'. This material may be edited as needed to align with site s13ecific characteristics.
* Any site s13ecific background or technical a1313endices that tile develo13ers belie 1 le would be useful in e>,13laining or using elements of tile emergency classification scheme. ,ti, Basis section should not contain information that could modif;' tile meaning or intent of tile associated IC or EAL. Suell information should be inoor13orated within tile IC or EAL statements , or as an EAL J>Jote. Information in tile Basis should only clarify and inform decision making for af1 emergency classification. Basis information should be readily a 1 1ailable to be referenced , if necessary , by tile Emergency Director.
for e>rnm13le , a co13y of tile basis document could be maintained in tile a1313ro13riate emergency res13onse facilities. Because tile information in a basis document can affect emergency classification decision making (e.g., tile Emergency Director refers to it during an event), tile NRG staff e>(13ects that changes to tile basis document will be evaluated in accordance with tile 13rovisions of 10 CfR 50.54(q).
4.-+4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA As reflected in the generic guidance,Some of the criteria/values used in several EALs and fission product barrier thresholds may be drawn from a 13lant'sPBNP
's AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments. Develo13ers should verify that aA ppropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to determine if an evaluation pursuant to 10 CPR 50.54(q) is required.
28 
 
===1.8 DEYELOPER===
 
AND USER FEE9BACI( l>lel 99 0 I (Re,*isieA 6) l>levember 2012 QuestioAs or commeAts coAcerniAg the material iA this documeAt may be directed to the tffil BmergeAcy PreparedAess staff, tffil BAL task force members or submitted to the BmergeAcy PreparedAess FrequeAtly Asked QuestioAs process. 29 5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS
 
===5.1 GENERAL===
CONSIDERATIONS l>Je l 99 0 I (Re~*i s ieA 6) l>levember 2012 When making an emergency cla s sification , the E mergency Dir e ctor must cons id er all information having a bear in g on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (E AL) plu s the associ a ted Operating Mode Applicability , Notes and the informing Basi s information.
In th e Recognition Categor y F matr i ces , EALs are referred to as Fi s sion Product Barrier Thre s holds; the thresholds serve the same function a s an EAL. NRC regulations require the li censee to estab li s h and maintain the capa bili ty to assess , class i fy , and declare an emergenc y condition w ithin 15 minutes after t h e availabi lit y of indications to plant operators that an emergency action level ha s been exceeded and to promptly declare the emergenc y condition as soon as possible follow in g identification of the appropriate emergency clas s ificat i on level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01 , Interim Staff Guidance , Emerg e n cy Planning f or Nucl e ar Pow e r Plant s. All emergency c l assificatio n assessments shou ld be based up on va l id ind ications , reports or conditions.
A valid indication, report , or condition, is one that has been verified through appropriate means s u ch that there is n o d oubt regarding the indic ator's operabi li ty , the condition's existence , or the report's accuracy.
For example , va lidati on could be accomp l ished through an instrument channel check , re s ponse on related or redundant indicators , or direct observation by plant personnel.
The va lid ation of indications shou ld be comp l eted in a manner that supports timel y emergency declaration. For ICs and EALs that have a stipu l ated time duration (e.g., 15 minutes , 30 minutes , etc.), the Emergency Director sho uld not wait until the applicab l e time has e lap sed , but should declare the event as soon as it is determined that the condition ha s exceeded , or will likely exceed , the applica ble time. If an ongoing radiolog i cal release i s detected and the release start time i s unknown , it sho uld b e assumed that the release durati on spec ifi ed in the IC/EAL has been excee ded , abse nt data to the contrary. For EAL thresholds that specify a duration of the off-normal condition.
the NRC expects that the emergency declaration process run concurrently with the specified threshold duration.
Once the off-normal condition has existed for the duration specified in the EAL, no further effort on this declaration is necessary-the EAL has been exceeded.
Consider as an example. the EAL " fire which is not extinguished within 15 minutes of detection." On receipt of a fire alarm. the plant fire brigade is dispatched to the scene to begin fire suppression efforts. If the fire brigade reports that the fire can be extinguished before the specified duration, the emergency declaration is placed on hold while firefighting activities continue.
If the fire brigade is successful in extinguishing the fire within the specified duration from detection.
no emergency declaration is warranted based on that EAL. 30 1>1e l 99 0 I (RevisieR
: 6) }18Y8ffi08r 20 J 2
* If the fire is still burn in g after the specified duration h as elapsed. the EAL i s exceeded.
no further assessment is necessary, and the emergency declaration would be made promptly. As used here. " promptly" means at the first available opportunity (e.g .. if the Shift Manager i s receiving an u pdate from the fire brigade at the 15-minute mark. it i s expected that the declaration w i ll occ u r as the next act i on after the ca ll ends).
* If. for example. the fire brigade notifie s the shift supervision 5 minutes after detection that the br i gade itself cannot ex tinguish the fire such that the EAL will be met imminently and cannot be avo i ded. the NRC would not consider it a vio l ation of the licensee's emerge n cy p l an to declare the event before the EAL is met (e.g .. the 15-minute duration has elap se d). While a prompt declaration wo u ld be beneficial to public health and safety and is encouraged, it is not requ i red by regulation.
* In all of the above. the fire duration is measured from the time the alarm, indication.
or report was first rece i ved by the p l ant operators. Validation or confirmation establishes that the fire started as early a s the time of the a l arm , i nd i cation. or report. A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that 1) the activit y proceeds as planned and 2) the plant remains w ithin the limits imposed by the operating license. Such activities includ e planned work to test , manipulate , repa ir , maintain or modify a system or component.
In these cases , the contro l s associated with the planning , preparation and execution of the work will ensure that comp li ance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 ~CFR 50.72. The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose a s sessments , chemistry samp lin g , RCS leak rate calculation , etc.); the EAL and/or the associated basis discussion wi ll identify the necessary analysis.
In these cases , the 15-minute declaration period starts with the availab ili ty of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects li censees to estab li sh the capabi lity to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).
Wh il e the EALs have been developed to address a full spectrum of possible events and conditions which may warra nt emerge nc y classification , a provision for clas s ification based on operator/management exper i ence and judgment is still necessary. The NEI 99 O+ Th i s scheme provides the Emergency Director with the ability to c l assify even t s and conditions based upon judgment using EALs that are consistent with the Emergency Classificat i on Leve l (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated into the Fi s sion Product Barrier Tab l es; judgment may be used to determine the status of a fission product barrier. 31 
 
===5.2 CLASSIFICATION===
 
METHODOLOGY Nel 99 0 I (Re\*isieR 6) }l9Y0FR00f
?Q 12 To make an emergency classification , the u s er will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or e x ceeded. The ev a luation of an E AL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded , then the IC is considered met and the a s sociated ECL i s declared in accordance with plant procedures.
When assessing an E AL that specifies a time duration for the off-normal condition , the " clock" for the EAL time duration runs concurrently with the emergency classification process " clock." For a full discussion of this timing requirement , refer to NSIR/DPRISG-01. 5 .3 CLASSIFJCA TION OF MULTIPLE EVENTS AND CONDITIONS When multiple emergency events or conditions are present , the user will identify all met or exceeded EALs. The highest applicable E CL identified durin g this review is declared. For example:
* If an Alert EAL and a Site Area Emergency EAL are met , whether at one unit or at two different units , a Site Area Emergency should be declared.
There is no " additive" effect from multiple E ALs meeting the same ECL. For example:
* If two Alert EALs are met , whether at one unit or at two different unit s, an Alert should be declared.
Related guidance concerning cla s sification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02 , Clarifi c ation of N R C Guidan ce for Em e rg e ncy Notifications During Quickly C hangin g Events. 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time that an event or condition occurred , and prior to any plant or operator response , is the mode that determines whether or not an IC is applicable.
If an event or condition occurs , and results in a mode change before the emergency is declared , the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached , any new event or condition , not related to the original event or condition , requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.
For events that occur in Cold Shutdown or Refueling , escalation i s via EALs that are applicable in the Cold Shutdown or Refuelin g modes , even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response.
In particular , the fission product 32 
}IEI 99 0 I (Revision
: 6) }loveFneer 2012 barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 5.5 CLASSIFICATION OF IMMINENT CONDITIONS Although EALs provide specific thresholds , the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If , in the judgment of the Emergency Director , meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels , this approach is particularly important at the higher emergency classification levels since it provides additional time for implementati o n of protective measures. 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING An ECL may be downgraded when the event or condition that meets the highest I C and EAL no longer exists, and other site-specific downgrading requiremen ts are met. If downgrading the ECL is deemed appropriate , the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated.
The following approach to downgrading or terminating an ECL is recommended.
ECL Action When Condition No Longer Exists Unusual Event Terminate the emergency in accordance with plant procedures.
Alert Downgrade or terminate the emergency in accordance with plant procedures.
Site Area Emergency with no Downgrade or terminate the emergency in long-term plant damage accordance with plant procedures.
Site Area Emergency with Terminate the emergency and enter recovery in long-term plant damage accordance with plant procedures.
General Emergency Terminate the emergency and enter recovery in accordance with plant procedures. As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02. 33 
 
===5.7 CLASSIFICATION===
 
OF SHORT-LIVED EVENTS l>lEI 99 0 I (Revisiefl
: 6) l>levelf!ser
?Q I 2 As discussed in Section 3 .2, event-based I Cs and EALs define a variety of specific occurrences that have potential or actual safety significance.
B y their nature, some of these event s may be short-lived and , thus , over before the emergency classification assessment can be completed.
If an event occurs that meets or exceeds an EAL , the associated ECL must be declared regardless of its continued presence at the time of declaration.
Examples of such events include a failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip or an earthquake.
 
===5.8 CLASSIFICATION===
 
OF TRANSIENT CONDITIONS Many of the ICs and/or EALs contained in this document employ time-based criteria.
These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified , it is recognized that some transient conditions may cause an EAL to be met for a brief period oftime (e.g., a few seconds to a few minutes).
The following guidance should be applied to the classification of these conditions.
EAL momentarily met during expected plant response -In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected , and operator actions are performed in accordance with procedures.
EAL momentarily met but the condition is corrected prior to an emergency declaration
-If an operator takes prompt manual action to address a condition , and the action is successful in correcting the condition prior to the emergency declaration , then the applicable EAL is not considered met and the associated emergency declaration is not required.
For illustrative purposes , consider the following example.
An ATWS occurs and the auxiliary feedwater system fails to automatically start. Steam generator levels rapidly decrease and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad and RCS barriers).
If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal condition prior to an emergency declaration , then the classification should be based on the A TWS only. It is important to stress that the 15-minute emergency classification assessment period is not a " grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event; emergency classification assessments must be deliberate and timely , with no undue delays. The provision discussed above addresses only those rapidly evolving situations where an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration.
34 l>lel 99 0 1 (ReYisieR
: 6) l>leve ffi eer 20 1 2 This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION In some cases , an EAL may be met but the emergenc y classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL , but no emergency was declare d , and the event or condition no longer exists at the time of discovery.
Thi s may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process. In these cases , no emergency declaration is warranted; however , the guidance contained in NUREG-1022 is applicable. Specifically , the event should be reported to the NRC in accordance with 10 CFR ,&sect;-50.72 within one hour of the discovery of the undeclared event or condition.
The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.
5 .10 RETRACTION OF AN EMERGENCY DECLARATION Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022. 35 l>le l 99 0 I (ReYisieA
: 6) Deee!flber 20 I 0 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS Table AR 1: Reeegeitien CategeFY "AR" Initiating Cenditiee MetFix UNU8Ui-..L EVENT AUlRUl Release of gaseous or liquid radioactivity greater than 2 times the (site specific effluent release controlling document)ODCM limits for 69 minutes or longer. Op. },fefics:
All i4 .. U2RU2 U}JPLA1'l}J ED loss of 1 t11ater level above irradiated fuel. Op. !Jerics: All ALERT .t..AlRAl Release of gaseous or liquid radioactivity resulting in offsite dose greater than IQ mrem TEDE or 59 mrem thyroid COE. Op. !,{erics:
All Significant lowering ofv,*ater level above , or damage to, irradiated fuel. Op. },{erics:
All 8ITEAREA EMERGENCY A81R81 Release of gaseous radioactivity resulting in offsite dose greater than 199 mrem TEDE or 599 mrem thyroid COE. Op. },feric s: All Spent fuel pool Je .. *el at (site specific Level 3 description)iQ ft 8 in (Level 3}. , ... ..,i\3B:fd
-Radiation levels that impede access to equipment necessary for normal plant operations , cooldown or shutdown.
Op. Maries: All 36 GENERA.L EMERGENCY AGlRGl Release of gaseous radioactivity re s ulting in offsite dose greater than 1 , 999 mrem TEDE or 5 , 999 mrem thyroid COE. Op. }.{erics:
Spent fuel pool level cannot be restored to at least (site specific Level 3 description)49 ft 8 in (Level 3) for 69 minutes or longer. :AU -
ECL: Notification of Unusual EventUnusual Event }le i 99 0 I (Re~*isieR 6) Nevember 3012 AU1RU1 Initiating Condition:
Release of gaseous or liquid radioactivity greater than 2 times the specific effluent release controlling document)ODCM l imits for 60 minutes or longer. Operating Mode Applicability:
All Emergency Action Levels: Example Emergeney ,A .. etion Levels: (1 or 2 or 3) Notes:
* The Emergency Director should declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the re l ease start time is unknown , assume that the release duration has exceeded 60 minutes.
* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path , then the effluent monitor reading is no lon ger valid for classification purposes.
Reading on ANY of the following effluent radiation monitor~ greater than the reading shown for 2 times the (site specific effluent release controlling document) limits for 60 minutes or longer: Monitor Reading 1 (2)-RE-307 CTMNT Purge Exhaust Mid-Range Gas l .4E-2 !!Ci/cc with onlv containment gurge in ogeration 2-RE-305 CTMNT Purge Exhaust Low Range Gas 9.4E-3 !!Ci/cc with both gur g e and GS building ventil a tion in ogeration 2-RE-307 CTMNT Purge Exhaust Mid-Range Gas 9.4E-3 !!Ci/cc with both gurge and GS building ventil a tion in ogeration 2-RE-307 CTMNT Purge Exhaust Mid-Range Gas 2.8E-2 !!Ci/cc with only GS building v entilation in ogeration 2-RE-307 CTMNT Purge Exhaust Mid-Range Gas I.OE+ 1 !!Ci/cc with onlv for c ed vent o f containment 2-RE-309 CTMNT Purge Exhaust High Range Gas LOE+ l !!Ci/cc with only for ce d vent o f c o ntainment RE-315 AB Exhaust Low Range Gas 5.4E-3 gCi/cc RE-317 AB Exhaust Mid-Range Gas 5.4E-3 gCi/cc RE-325 Drumming Area Exhaust Low Range Gas 8.4E-3 gCi/cc RE-327 Drumming Area Exhaust Mid-Range Gas 8.4E-3 gCi/cc 1 (2)-RE-229 Service Water Overboard 2.3E-3 !!Ci/cc 37 l>le l 99 O I (Re, 1 isieR 6) l>l e, 1 em0er ?Q 12 R 1.2+/- (site specific monitor list and threshold values corresponding to 2 times the controlling document limits) Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer. ~3 Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the (site specific effluent release controlling document)ODCM limits for 60 minutes or longer. 38 Definitions:
Basis: NEI 99 0 I (RevisioA
: 6) l>loveffieer 2012 This IC addresses a potential decrease in the level of safety of the plant as indicated by a level radiological release that exceeds regulatory commitments for an extended period oftime (e.g., an uncontrolled release).
It includes any gaseous or liquid radiological release, monitored or un--monitored, including those for which a radioactivity discharge permit is normally prepared.
l'Juc l ear pov, , er plantsPBNP incorporates design features intended to control the release of radioactive effluents to the environment.
Further , there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases.
The occurrence of an extended , uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. __ The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Classification based on effluent monitor readings assumes that a release path t o the environment is established. Jf the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path , then the effluent monitor reading is no longer valid for classification purposes. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. EAL RU 1.1 --This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.
EAL RUI .2 --This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. __ This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas). EAL RU 1.3 --This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems , etc.). Escalation of the emergency classification level would be via IC AA-l-RA I. Developer Notes: The "s i te specific effluent release controlling document" is the Radio l ogical Effluent Technical Specifications (RETS) or, for plants that have i mplemented Generic Letter 89 01 \--the + hnpl-emen.'elien ef Pregremme.'ic Centrals fer Refi.ielegicel Effluent Technical Spccificetiens in /he Adminislre/ive Gen Ir els &clien 9} /he Teclmicel Spccificeliens end the Rclecelien 9} Preccthwel Dc,'eils ~fRET8 le lhc Ojfsilc Dase Celc1.lelien klam,1el er .'e .'he Precess Ce1ilrel Pregram 39 01 (RevisieA
: 6) }JI;J 99 ber ?Q 12 }leveFA -
:g,1 L using est to ineluae an l ' *
* A licensee may requ case by case basts. I t Teehnical S13eeificat1ons.
'a',' .. , j]I be consiaerea on a "'the "'*an--r-It . a"'"'rov 1,Y * "bs !lie seope o, P. . system ,ese , , 7P . d d .* !lie geRef!e en-.
* dose p~eot,o* re RO! **elaee 1 , oot Be real '""'o *..,ere, fflOoito,iog systeoa11i:, do tliese mooi!oFS "'~ e of tlie plaot ---ttlnttjElmic=ations frm: ~,:e;~is ca13ability.
For! tli~s:naipffle:it, or witliio _!lie::: .. , or o!lier '"=~ooi:ffiS~eetsHEl:kot-nn<o~t:1t1.,__.,a, ie"el asp OR " d l,y eovireo Mafty lieeose--. t ioed !o !lie soffie ****. "'"Y 1,e iollaeoee-
-oitoriog systeffi; eootrolletl
**d ffiOIR a s lo odditioo , read mg~' b esiog o periffietor me Teelioieal Spee1fiea~:~
;eq*est to ioel*de "" a:,;, factors. A licensee *seres on a case by case . 'al '"ill be CORSI a1313ro1v " . nt 6,ttributes:
BCL Ass1gnme 3.1.1.B 41 ECL: Notification of Unusual EventUnusual Event l>Jel 99 O I (Re*,*isieA 6) l>JeYeFAl:Jer 2012 AU2RU2 Initiating Condition:
UNPLANNED loss of water level above irradiated fuel. Operating Mode Applicability:
All i E .. mple Emergency Action Levels: + a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
* Spent fuel pool low water level alarm
* Visual observation(site specific level indications).
AND b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
* (site specific list of area radiation monitors)RE-105 SFP Area Low Range Radiation Monitor
* RE-135 SFP Area High Range Radiation Monitor
* 1{2):RE-102 El. 66' CONTAINMENT Low Range Monitor Definitions:
UNPLANNED:
A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal. Basis: This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. A water level decrease will be primarily determined by indications from available level instrumentation.
The low level alarm is actuated by LC-634, SFP Level Indicator at 62 ft. 8 in. based on maintaining at least 6 ft. of water on a withdrawn fuel assembly.
_Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available).
A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.
42 l>lel 99 01 (RevisieH
: 6) J>levemeer 2012 The effects of planned evolutions should be considered.
For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly.
Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC AMRA2. Devel0f)er N0tes: The "s ite specific level indications" are those indications that may be used to monitor 1 n'ater level in the various portions of the REFUELING PATH'NAY.
Specify the mode applicabilit)'
of a particular indication if it is not available in all modes. The "s ite specific list of area radiation monitors" should contain those area radiation monitors that would be e1,pected to have increased readings following a decrease in v,ater level in the site specific REFUELING PATHWAY. In cases *where a radiation monitor(s) is not available or v,ould not provide a useful indication, consideration should be given to including alternate indications such as UNPLA1'l1'IBD changes in tank and/or sump levels. Development of the EALs should consider the availability and limitations of mode dependent , or other controlled but temporary, radiation monitors.
Specify the mode applicabilit)
' of a particular monitor if it is not available in all modes. EGL Assignment Attributes:
3 .1.1./\ and 3 .1.1.B 43 l'lel 99 0 I (Revision
: 6) l'loveFnaer 2012 AA1RA1 ECL: Alert Initiating Condition:
Release of gaseous or liquid radioactivity resultin g in offsite dose greater than IO mrem TEDE or 50 mrem thyroid CDE. Operating Mode Applicability:
All Emergency Action Levels: Example Emergeney Aetion LeYels: (1 or 2 or 3 or 4) Notes:
* The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded , or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown , assume that the release duration has exceeded 15 minutes.
* If the effluent flow past an effluent mon i tor is known to have stopped due to actions to isolate the release path , then the effluent monitor reading is no longer valid for classification I
* Rhl l purpo ses. The pre-calculated effluent monitor values presented in EAL RA 1.1 s hould illl!y_be u se d for emergency classification assessments until the re s ult s from a dose assessment u s ing actual meteorology are available.
Readin g on ANY of the following radiation monitor s greater than the reading shown fo r 15 minutes or longer: Monitor Reading 1(2)-RE-307 CTMNT Purge Exhaust Mid-Range Gas 6.0E+O !!Ci/cc with onl:r containment 12urge in 012eration 1 (2)-RE-309 CTMNT Purge Exhaust High Range Gas 6.0E+O b!Ci/cc with onl:r containment 12urge in 012eration 2-RE-307 CTMNT Purge Exhaust Mid-Range Gas 4.0E+O !!Ci/cc with both 12urge and GS building ventilation in 012eration 2-RE-309 CTMNT Purge Exhaust High Range Gas 4.0E+O !!Ci/cc with both 12urge and GS building ventilation in 012eration 2-RE-309 CTMNT Purge Exhaust High Range Gas l.2E+ 1 !!Ci/cc with onl:r GS building ventilation in 012eration RE-317 AB Exhaust Mid-Range Gas l .OE+O b!C i/cc RE-319 AB Exhaust High Range Gas l .OE+O b!Ci/cc RE-327 Drumming Area Exhaust Mid-Range Gas l.6E+O b!Cilcc 44 
.fili.L 2 (site specific moAitor list aAd threshold values) l>IEI 99 0 I (ReYisieA
: 6) l>levefl'!ber 2012 Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site specific dose receptor poiAt)SITE BOUNDARY.
Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond specific dose receptor poiAt)the SITE BOUNDARY for one hour of exposure. Field survey results indicate EITHER of the following at or beyond (site specific dose receptor poiAt)the SITE BOUNDARY:
* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.
45 Definitions:
}/el 99 O I (Re\*i s i e A 6) }le velflb e r 2012 SITE BOUNDARY:
That lin e b e yond which the land is neither own e d. nor leased. nor other w i s e controlled by the licensee. Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
This IC is modified by a note that EAL RA 1.1 is only assessed for emergency classificat i on until a qualified dose assessor is performing as s essments using dose projection software incorporating actual meteorological data and current radiological conditions.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1 % of the EPA PAG of 1 , 000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established.
_If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
For EAL RAl.3. there are no s ite-specifi c liquid radiation monitors capable of monitoring liquid effluent releases at the classification threshold for this EAL because their detector operating range is exceeded prior to reaching these levels. Entry into this EAL for a liquid radioactivity release will be based on.,. sampling initiated due to entry into E AL RUl. In practical terms. this means that entry into IC RUl will start sampling (per RMS Alarm Setpoint and Response Book) which will then allow detection of the s etpoint for RA 1. ---Escalation of the emergency classification level would be via IC A-8-l-RS 1. De1,*el0fJeF Nates: '.Vhile this IC may Hot be met ab s eHt challeHges to oHe or more fissioH product barriers , it provides classificatioH diversity aHd may be used to classify eveHts that *.vould Hot reach the same EGL based OH plaHt status or the fissioH product matri>, aloHe. for maHy of the DBAs analyzed in the Updated fiHal Safety Analysis Report , the discriminator will not be the Humber of fission product barriers challenged , but rather the amount of radioactivit)
' released to the eHvironment.
46 
?>l e i 99 0 I (ReYision 6) }JOYeffieer 2012 The EPA PAGs are eJ{pressed in terms of the sum of the effective dose equi*,alent (EDE) and the committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent (CDE). For the purpose of these IC/EALs, the dose quantity total effective dose equivalent (TEDE), as defined in 10 CFR &sect; 20, is used in lieu of" ... sum of EDE and CEDE. ... ". The EPA PAG guidance provides for the use of adult thyroid dose conversion factors; however, some states have decided to base protective actions on child th)'roid CDE. Nuclear pov,'er plant ICs/EALs need to be consistent 1,yith the protective action methodologies employed by the States ,,,,,ithin their EPZs. The thyroid CDE dose used in the IC and EALs should be adjusted as necessary to align with State protective action decision making criteria.
The "site specific monitor list and threshold values" should be determined with cons i deration of the following:
* Selection of the appropriate installed gaseous and liquid effluent monitors.
* The effluent monitor readings should correspond to a dose of 10 mrem TEDE or 50 mrem thyroid CDE at the "site specific dose receptor point" (consistent 1 Nith the calculation methodology employed) for one hour of eJ{posure.
* Monitor readings 1 Nill be calculated using a set of assumed meteorological data or atmospheric dispersion factors; the data or factors selected for use should be the same as those employed to calculate the monitor readings for ICs ASl and AGl. Acceptable sources of this information include , but are not limited to , the RETS/ODCM and values used in the site's emergency dose assessment methodology.
* The calculation of monitor readings 1 Nill also require use of an assumed release isotopic miJ{; the selected mb{ should be the same as that employed to calculate monitor readings for res ASl and AGL Acceptable sources of this information include, but are not limited to, the RETS/ODCM and Yalues used in the site's emergency dose assessment methodology.
* Depending upon the methodology used to calculate the EAL values, there may be overlap of some values between different ICs. Developers will need to address this overlap by adjusting these values in a manner that ensures a logical escalation in the EGL. The " site specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish between on site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan , and the procedural methodology used to determine offsite doses and ProtectiYe Action Recommendations. The variation in selected dose receptor points means there ma)' be some differences in the distance from the release point to the calculated dose point from site to site. Developers should research radiation monitor design documents or other information sources to ensure that 1) the EAL value being considered is within the usable response and display range of the instrument, and 2) there are no automatic features that may render the monitor reading invalid (e.g., an auto purge feature triggered at a particular indication level). It is recognized that the cond i tion described by this IC may result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. 1A those cases, EAL yalues should be determined with a margin sufficient to ensure that an accurate monitor reading is available.
For eJtample, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision nohvithstanding , if the est i mated/calcu l ated monitor reading is greater than apprmdmatel)'
110% of the highest accurate monitor reading, then developers may choose not to include the n'lonitor as an indication and 47 identify an alternate EAL threshold.
l>JE I 99 0 I (ReYisieA
: 6) l>l evember 2012 Although the IC references TEDE , field survey results are generally available only as a '\vhole body" dose rate. For this reason , the field survey EAL specifies a " closed *.vindow" survey reading. Indications from a real time dose projection system are not included in the generic EALs. Many licensees do not have this capability.
For those that do , the capability may not be within the scope of the plant Technical Specifications.
A licensee may reque s t to include an EAL using real time dose projection system result s; approval will be considered on a case b;* case basis. Indications from a perimeter monitoring system are not included in the generic EALs. Many licensees do not have this capability.
For those that do , these monitors may not be controlled and maintained to the same level as plant equipment, or within the scope of the plant Technical Specifications.
In addition , readings may be influenced by environmental or other factors. A licensee may request to include an EAL using a perimeter monitoring system; approval *.viii be considered on a case by case basis. EGL Assignment Attributes:
3.1.2.C 48 ECL: A l ert J>Je l 99 0 I (R e vi sie A 6) J>le v e mber 2 0 12 AA2RA2 Initiating Condition:
Significant l owering of water l eve l above , or damage to , irradiated fuel. Operating Mode App licability:
A ll Emergency Action Levels: E1rnmple Emergency Action Levels: (1 or 2 or 3) Uncovery of irradiated fuel in the REFUELING PATHWAY. Damage t o irradiated fuel re s ultin g in a rel ease of radioactivity from the fuel as indicated by a reading on ANY of the following radiation monitors greater than the value shownANY of the following radiation monitors:~
Monitor Reading RE-105 SFP Area Low Range Radiation Monitor 4 R/hr 1 (2)-RE-126 Containment High Radiation Monitor 7 R/hr 1(2)-RE-127 Containment High Radiation Monitor 7 R/hr 1 (2)-RE-128 Containment High Radiation Monitor 7 R/hr RA2.3 (site specific listing of radiation monitors , and the associated readings , setpoints andJor alarms) L wering of s pent fuel pool l evel to (site specific Level 2 value). [Sec Dcvclepcr
}'letes]49
-ft.Qin. Definitions:
REFUELING PATHWAY -The reactor refueling cavity. spent fuel pool and fuel transfer canal. Basis: This IC ad dr esses events that have ca u sed IMMINENT or actual damage to an irr adiate d fuel assembly , or a s i gnificant l ower in g of water le vel w ithin the spent fuel poo l (sec Dcvclepcr Nete-s}. These events present radiological safety cha ll enges to plant personnel and are precursors to a release of radioactivity to the e n vironme nt. As suc h , they represent an actual or potential substantial degrad at i o n of the level of safe t y of th e plant. This IC app li es to irradiated fuel that i s licensed for dry s t orage up to the poin t th at the load ed storage cask i s sealed. Once sea l e d , d amage to a l oade d cask caus in g l oss of the CONFINEMENT BOUNDARY is classified in accor dan ce w ith IC E-H Ul. Esca lati on of the emerge n cy wo uld b e based on e ith e r Recognition Category A-R or CI Cs. 49 EAL RA2.l Ne.I 99 0 I (RevisioA
: 6) l>lOY0ffiB0f 2012 This EAL esca l ates from ~RU2 in that the l oss of l evel, in the affected portion of the REFUELING PATHWAY , is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels , or other plant parameters.
Computational aids may also be used (e.g., a boil off et:tf&#xa5;e1. Classification of an event u sing this EAL shou ld be based on the totality of avai l able indications , reports , and observations. 50 l>JE I 99 0 I (Re v isie A 6) l>Jeve FAeer 2 0 I 2 While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible , readings should be considered in combination with other available indications of inventory loss. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. EAL RA 2.2 This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly.
A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damag i ng event (e.g., a fuel handling accident).
EAL RA2.3 Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability t o adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via I Cs AS-1--RS 1 or ~RS2(see A S 2 Devdf:Jper ,Vales). Developer Notes: For EAL #1 Depending upon the availabilit y and range o f instrumentation , this EAL may include specific readings indicative of fuel un c overy; consider water and radiation level reading s. Specif)' the mode applicability of a particular indication if it is not available in all modes. F or EAL #2 The " site specifi c li s ting of radiation monitor s, and the associated readings , setpoints and/or alarms" should contain those radiation monitors that could be used to identify damage t o an irradiated fuel assembly (e.g., confirmatory of a relea s e of fission product gases from irradiated fuel). F or EALs # 1 and #2 Developers should research radiation monitor design documents o r other information source s to ensure that 1) the E AL valu e being con s idered i s within the usable respon s e and display range of the in s trument , and 2) there are n o automatic feature s that may render the monitor reading invalid (e.g., an auto pur g e feature tri gg ered at a particular indication level). It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range o f the installed radiation monitor. In those cases , EAL 51 N E I 99 01 (Re\*i s ion 6) ~l oYeffiber 2012 values should be determined with a margin sufficient to ensure that an accurate monitor reading is available.
For example , an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than apprmdmately 110% of the highest accurate monitor reading , then developer s may choose not to include the monitor as an indication and identify an alternate EAL threshold. To further promote accurate classification , developers should consider if s ome combination of monitors could be specified in the B AL to build in an appropriate level of corroboration between monitor readings into the classification assessment.
Development of the EALs should also con s ider the availability and limitations of mode dependent , or other controlled but temporary , radiation monitors.
Specify the mode applicability of a particular monitor i f it i s not available in all modes. For EAL #3 In accordan e e v,*ith the discussion in Section 1.4 , }JRG Order EA 12 051 , it is recommended that this EAL be implemented 1 Nhen the enhanced spent fuel pool level instrumentation is available fur use. The " site specific Level 2 value" is usually the spent fuel pool level that i s adequate to provide substantial radiation shielding fur a person standing on the spent fuel pool operating deck. Thi s site s pe c ific level is determined in accordance with NRG Order EA 12 051 and }ffil 12 02 , and applicable owner's group guidance.
Developers s hould modify the EAL andlor Basis section to reflect any site specific constraint s or limitations assoeiated with the design or operation of instrumentation used to determine the Level 2 value. EGL Assignment Attr i butes: 3 .1.2.B and 3 .1.2.G 52 ECL: Alert }l e i 99 0 I (Revisien 6) }Jevember 2012 AA3RA3 Initiating Condition:
Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability:
All Emergency Action Levels: Examf)le Emergeney
} .. etien Levels: (1 or 2) Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
3.1 Dose rate greater than 15 mR/hr in ANY of the following areas:
* Control Room (RE-101) ._Central Alarm Station AND * (other site specific areas/rooms)Secondary Alarm Station (by survey) 3.2 An UNPLANNED event results in radiation levels that prohibit or impede access to any of the following plant rooms or areas: Area Mode Ul VCT Area 3/4/5 U2 VCT Area 3 /4 I 5 Ul Primarv Samole area 3 U2 Primarv Samole area 3 CCWHXRoom 4/5 C-59 area 3/4/5 Pioewav 2. 8 ft. Elev. 3/4 Pioewav 3. 8 ft. Elev. 3/4 l/2B32 MCC Area 4 (site specific list of plant rooms or areas 1 with entry related mode applicability identified)
Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plan t operation, or to perform a normal plant cooldown and shutdown.
As such , it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should 53 
~JEI 99 0 I (RevisieA
: 6) ~le~*em~er ?Q I 2 consider the cause of the increased radiation levels and determine if another IC may be applicable.
54 1>Je l 99 O 1 (Re~*i s i e A 6) Ne vember 2 0 12 For EAL RA3 .2 , an Alert declaration is warranted if entry into the affected room/area is , or may be , procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entr y is actually necessary at the time of the increased radiation lev e ls. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area ( e.g., insta ll ing temporary shielding, requiring use of non-routine protective equipment , requesting an extension in dose limits beyond normal administrative limits). An emergency declaration is not warranted if any of the following conditions apply.
* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example , the plant is in Mode 1 when the radiation increase occurs , and the procedures used for normal operation , cooldown and shutdown do not require entry into the affected room until Mode 4.
* The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibi l ity of a room or area ( e.g., radiography , spent filter or resin transfer, etc.).
* The action for which room/area entry is required is of an administrative or record keeping nature (e.g., norma l rounds or routine inspections).
* The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action. The list of plant rooms or area s in EAL RA3.2 was generated from a step-by-step r e view of OP-3A, 3B, 3C. 50. and 7 A. Escalation of the emergency classification level would be via Recognition Category AR , C or F ICs. Develeper Netes: EAL#l The value of 15mR/hr is derived from the GDC 19 value of 5 rem in 30 da y s with adjustment for e~(pected occupancy times. The " other site specific areas/rooms" should include any areas or rooms requiring cont i nuous occupancy to maintain normal p l ant operation , or to perform a normal cooldovm and shutdov,rn.
EAL#2 The " site spec i fic li s t of plant rooms or areas with entry re l ated mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal p l ant operation , cooldovm and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed. (e.g., an action to addres s an off normal or emergency condition such as emergency repairs , correct i ve measures or emergency operations). In addition, the list should specify the p l ant mode(s) during which entry would be required for each room or area. 55 NEI 99 01 (Re~*isieR 6) }levemeer 2012 The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
If the equipment in the listed room or area *uas already inoperable, or out of service , before the event occurred , then no emergenc)'
should be deelared since the event will have no adverse impact beyond that already allovred by Technical Specifications at the time of the event. Rooms and areas listed in EAL #1 do not need to be included in EAL #2 , includ i ng the Control Room. EGL Assignment Attributes:
3.1.2.C 56 ECL: Site Area Emergency
~IE I 99 0 I (Revi s i e n 6) ~l evember 2012 AS1R51 Initiating Condition:
Release of gaseous radioactivity resulting in off site dose greater than 100 mrem TEDE or 500 mrem thyroid CDE. Operating Mode Applicability:
All Emergency Action Levels: Example Emergeney A.etion LeYels: (1 or 2 or 3) Notes:
* The Emergency Director should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded , or wi ll l ikely be exceeded.
* If an ongoing release is detected and the release start time is unknown , assume that the release duration has exceeded 15 minutes.
* If the effluent flow past an effluent monitor is known to have stopped due to actions t o isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
I
* The pre-calculated effluent monitor values presented in EAL RS 1.1 should_Qfily be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. fil l Reading on ANY of the following radiation monitor s greater than the reading shown for 15 minutes or longer: Monitor Reading 1 (2)-RE-307 CTMNT Purge Exhaust Mid-Range Gas 6.0E+ I !!C i/cc with onlv cont a inment gurge in ogeration I (2)-RE-309 CTMNT Purge Exhaust High Range Gas 6.0 E+ 1 b!:Ci/cc with onl v c o nt a inm e nt gurg e in oger a ti o n 2-RE-307 CTMNT Purge Exhaust M i d-Range Gas 4.0E+ I !!Ci/cc with both gur ge and GS building v e ntil a tion i n og e ration 2-RE-309 CTMNT Purge Exhaust H i gh Range Gas 4.0E+ I !!Ci/cc w ith both gurg e a nd G S building ve nt i l a tion in og e r a tion 2-RE-309 CTMNT Purge Exhaust High Range Ga s l.2E+2 !!Ci/cc w ith onl v G S building ve ntil a tion in o g e rat i on RE-317 AB Exhaust Mid-Range Gas 1.0E+l b!:Ci/cc RE-319 AB Exhaust High Range Gas I.O E+ I b!:Ci/c c RE-327 Drumming Area Exhaust Mid-Range Gas 1.6 E+ 1 b!:Ci/cc 57 
~IE I 99 0 I (R e vi s i e A 6) ~le vemeer ?Q 12 Rfil_,----1(.,.s+i:it@,.e
...,S.Ap~ec~if'J'filf'C.-Jffi~OnR-i~toRirF--lwi S'-"'t~al-J:ln,f:ld--i:tCR-1,ure~s'R-h~o Jucd....,,1,4,a:i.1-Jw,ue~s.4) 2 R 1.3 Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond (site specific dose receptor point)the SIT E BOUNDAR Y. Field survey results indicate EITHER of the following at or beyond (site specific dose re c eptor point)the SIT E BOUNDARY:
* Closed window dose rates greater than 100 mR/hr expected to continue for 60_ minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.
58 Definitions:
1'JJ;I 99 0 1 (R e~*i sie A 6) 1'JS Y 0 ffi0er 2 0 12 SITE BOUNDARY:
That line beyond which the l a nd i s n e ither o w n e d , nor lea s ed. nor o t herwi s e controlled by the licen s e e. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. This IC i s modified by a note that EAL RS 1.1 i s only a ss e ss ed for em e rgency cla ss ification until a qualified do se a ssess or is performing assessment s using dose proj e ction software incorporating actual meteorologi c al data and current radiologic a l c onditions.
Radiological effluent EALs are also includ ed to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. _The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectru m of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established.
_ If the effluent flow past an effluent monitor is known to have s to pped due to actions to isolate the release path , then the effluent monitor reading is no longer valid for classification purposes.
Esca lati on of the emergency classification l eve l would be via IC AG+RG I. De11el0f)eF Nates: While thi s IC may not be met ab s ent challen g e s to multiple fi s sion produ c t barriers , it proYide s classification diver s it)* and may be used to classify events that 1 would not reach the same EGL based on plant s tatus or the fission product matrix alone. For many of the DBAs analyz:ed in the Updated Final Safety Analysis Report , the discriminator will not be the number of fission product barrier s challenged , but rather the amount of radioa c tivity relea s ed to the env i ronment. The EPA PAG s are e 1 cpressed in terms of the sum of the effective dose equivalent (ED E) and the committed effective dose equivalent (CED E), or as the thyroid committed do s e equivalent (CO E). Fo r the purpo s e of these lC/E ALs , the dose quantit y t o tal effective d o se equi 1 ralent (TED E), a s defined in 10 CFR &sect; 20 , i s u s ed in lieu of" ... s um. of ED E and C E DE .... ". The EPA PAG guidance provides for the use o f adult thyroid dose conversion factors; however , s ome state s have decided to base protective actions on child thyroid COE. *Nuclear power plant ICs/E AL s need to be consi s tent w i th the protective action methodologie s employed 59 
}le i 99 0 I (Revi s ieR 6) }J e~*effieer 2012 by the States *within the i r EPZs. The th)'roid COE dose used in the IC and EALs shou l d be adjusted as necessary to align with State protective action decision making criteria.
The " site specific monitor list and threshold values" should be determined 1+vith consideration of the following:
* Selection of the appropriate installed gaseous effluent monitors.
* The effluent monitor readings should correspond to a dose of 100 mrem TEOE or 500 mrem thyroid COE at the " site specific dose receptor point" (consistent with the calculation methodo l ogy employed) for one hour ofe>tposure.
* Monitor readings will be calculated using a set of assumed meteorological data or atmospheric dispersion factors; the data or factors selected for use should be the same as those employed to calculate the monitor readings for ICs AAl and AG 1. Acceptab l e sources of this information include , but are not limited to, the RETS/ODCM and values used in the site's emergency dose assessment methodology.
* The calculation of monitor readings will also require use ofan assumed release isotopic mi>{; the selected mi>t should be the same as that employed to calcu l ate monitor readings for lCs A,A.l and AGl. Acceptable sources of this information include, but are not limited to , the RETS/OOCM and Yalues used in the site's emergency dose assessment methodology.
* Depending upon the methodology used to calculate the EAL values , there may be overlap of some ,,alues between different ICs. Developers will need to address this overlap by adjusting these values in a manner that ensures a logical escalation in the EGL. The " site specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish between on site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan , and the procedural methodo l ogy used to determine offsite doses and Protective Action Recommendations. The variation in se l ected dose receptor points means there may be some differences in the distance from the release poiat to the calculated dose point from site to site. Developers should research radiation moaitor design documents or other information sources to ensure that 1) the EAL value being considered is 'n'ithin the usable response and display range of the instrument , and 2) there are no automatic features that may render the monitor reading invalid (e.g., an auto purge feature triggered at a particular ind i catioa level). It is recognized that the coadition described by this IC may result in a radiolog i cal effluent value beyond the operating or display range of the installed effluent monitor. In those eases , EAL values shou l d be determined with a margin sufficient to ensure that an accurate monitor reading is available. For e>rnmple , an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than apprmdmately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication aad identify an alternate EAL threshold. Although the IC references TEDE , field survey results are generally available only as a "whole body" dose rate. For this reason , the field survey EAL specifies a " closed window" survey reading. Indications from a real time dose projection system are not included in the generic EALs. 60 
}le i 99 01 (Re~'isieA 6) }l eYember 2012 Many lioensees Elo not ha,e th!S oapab~*~:
f= A lieensee may re~uest to ineluElo an E;:1' .""ng the scope of the plaHt !eehmeal Spee: ~: IO:o~al will be coHsidered OH a ease by case as1s. "'8itilfle Elose p<<ajeel1en systolfl rosu ts , pp . . E' bs . . . s *stern are not included m the generic n . .. , r those that do , the capability may not be ,\withi~ Indications from a perimeter m?~1tonng ) se that do , these monitors ma)' not be Many lioeesoes Elo not hiwe this eapab,/''.Y*, ::;:!'.:, e~uiplflent , er within tho soepe t''"::'""'
controlled aHd maintained to the ~~me eve. a , be influenced by environrnenta or o er Technical Specifications.
In add1t1?n , re;dmg~ L) u s ing a perimeter monitoring S)'Stem; . , reEiuest to mclu e an n factors. A licensee rn*d d a case by case basi s. approval will be cons, ere on EGL Assignment Attributes
: 3.1.3.C 61 J>JE~I 99 01 (Re\*isieR 6) Neveffiber 2012 AS2RS2 [See Develeper 1'letes] ECL: Site Area Emergency Initiating Condition:
Spent fuel pool level at (site specific Level 3 description)40 ft 8 in. Operating Mode Applicability:
All Example Emergency Action Levels: R 2.1 Lowering of spent fuel pool level to 40 ft. 8 in.(site specific Level 3 value). Definitions:
Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however , i t is included to provide classification diversity.
Escalation of the emergency classification level would be via IC AG-1-RG 1 or AQRG2. De-;el0per N0tes: In accordance with the discussion in Section 1.4, NRG Order EA 12 051 , it is recommended that this IC and EAL be implemented when the enhanced spent fuel pool le*;el instrumentation is available fur use. The "s ite specific Level 3 value" is usually that spent fuel pool level where fuel remains covered and actions to implement make up *water addition should no longer be deferred.
This site specific level is determined in accordance v,ith NRG Order EA 12 051 and NEI 12 02 , and applicable ovmer's group guidance.
Developers should modify the EAL and/or Basis section to reflect any site specific constraints or limitations associated with the design or operat ion of instrumentation used to determine the Level 3 value. EGL Assignment Attributes:
3 .1.3 .B 62 ECL: General Emergency J>JE I 99 0 I (ReYi s i e A 6) J>J0 Y8ffi08f 2 0 12 AR G1 Initiating Condition:
Relea se of gaseous radioactivity resulting in offsite dose greater t ha n 1 , 000 mrem TEDE or 5,000 mrem thyroid CDE. Operating Mode Applicability:
All Emergency Action Levels: LeYels: (1 or 2 or 3) Notes: Example Emergeney Aetion
* The Emergency Director should declare the Genera l Emergency promptly upon determining that the applicable time has been exceeded, or will lik ely be exceeded.
* If an ongoing release is detected and the release start time is unknown , assume that the release duration has exceeded 15 minutes.
* If the effluent flow past an effluent monitor is known to have stopped due t o actions to isolate the release path , then the effluent monitor reading is no l onger va lid for classification purposes.
I
* The pre-calculated effluent monitor values presented in EAL RG 1.1 shou ld...Q!l.[y be used for emergency classification assessments until the results from a dose assessment using actua l meteorology are available.
RQ.L l Reading on ANY of the following radiation monitor s greater than the reading shown for 15 minutes or longer: Monitor Reading 1(2)-RE-309 CTMNT Purge Exhaust High Range Gas 6.0 E+2 gCi/cc w ith onlv cont a inm e nt oure:e in oo e r a tion 2-RE-309 CTMNT Purge Exhaust High Range Gas 4.0 E+2 gCi/cc w ith both our e:e a nd G S buildin e: ve ntil a tion in oo erat ion 2-RE-309 CTMNT Purge Exhau s t High Range Gas l.2E+3 b!,Cilcc w ith onl v G S buildin e: ve n t il at ion in o o era ti o n RE-317 AB Exhaust Mid-Range Gas l.O E+2 gCi/cc RE-319 AB E x haust High Range Gas 1.0 E+2 gCi/cc (site specific moAitor li s t aAd thresh.old values) Dose assessment using actua l meteorology indicates doses greater than 1 , 000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond (site specific dose recept o r poiAt)the SIT E BOUNDARY. Field survey results indicate EITHER of the following at or be yo nd (site specific dose receptor poiAt)the SIT E BOUNDARY:
* C l osed window dose rates greater than 1,000 mR/hr expected to continue for 60_ minutes or longer. 63 1>1el 99 01 (Re\'isieA
: 6) }levember 2012
* Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.
64 Definitions:
1>1E;1 99 0 I (RevisieA
: 6) l>Je,*eff!ber 2012 SITE BOUNDARY:
That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.
Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equa l to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases.
Releases of this magnitude will require implementation of protective actions for the public. This IC is modified by a note that EAL RG 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions.
Radiological effluent EALs are also in cluded to provide a basis for classifying events and cond iti ons that cannot be readily or appropr iat ely classified on the basis of plant conditions alone. _The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose i s set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid CDE. C la ssification based on effluent monitor readings assumes that a release path to the environment is established. _I f the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Develef)eF Netes: The efflueAt ICs/EALs are iAcluded to provide a basis fur elassifyiAg eveAts that caAnot be readily classified OA the basis of plaAt coAditioAs aloAe. The iAclusioA of both t)'pes of ICs/EALs more full;' addresses the spectrum of possible eveAts aAd accideAts.
While this IC may Aot be met abseAt challeAges to multiple fissioA product barriers , it provides classificatioA di11ersity aAd may be used to classify eveAts that would AOt reach the same EGL based OA plaAt status or the fissioA product matri>t aloAe. For maAy of the DBAs aAalyzed iA the Updated FiAal Safety AAalysis Report , the discrimiAator will Aot be the Humber of fissioA product barriers challeAged , but rather the amouAt ofradioacti11it)'
released to the eAviroAmeAt.
The EPA PAGs are e>tpressed in terms of the sum of the effective dose equivaleAt (EDE) ood the committed effective dose equiYaleAt (CEDE), or as the thyroid committed dose equi11aleAt (CDE). For the purpose of these IC/EALs , the dose quaAtit)' total effecti11e dose equivaleAt (TEDE), as defined iA 10 CFR &sect; 20, is used in lieu of" ... sum of EDE and CEDE .... ". The EPA PAG guidance provides fur the use of adult th;'i'oid dose coA11ersion factors; however , some states have decided to base protective actions on child th;*roid CDE. Nuclear power plant ICs/EALs Reed to be consisteAt with the protective actioA methodologies employed 65 
~le l 99 01 (Re\*i s ieR 6) ~J 0Y6ffi06f
?Q 12 ey the States *Nithin their EPZs. The thyroid CDE dose used in the IC and eALs should ee adjusted as necessary to align with State protective action decision making criteria.
The " site specific monitor list and threshold values" should ee determined
', 1 ,ith consideration of the fol10 1 wing:
* Selection of the appropriate installed gaseous effluent monitors.
* The effluent monitor readings should correspond to a dose of 1 , 000 mrem TEDE or 5 , 000 mrem thyroid CDE at the " site specific dose receptor point" (consistent with the calculation methodology employed) for one hour ofeJtposure.
* Monitor readings will ee calculated using a set of assumed meteorological data or atmospheric dispersion factors; the data or factors selected for use should ee the same as those employed to calculate the monitor readings for ICs AAl and ASl. Aeeeptaele sources of this information include , eut are not limited to , the RETS/ODCM and values used in the site's emergenC)'
dose assessment methodology.
* The calculation of monitor readings will also require use of an assumed release isotopic mix; the selected mix should ee the same as that employed to calculate monitor readings for lCs /Vt.I and AS 1. Acceptaele sources of this information include, but are not limited to , the RETS/ODCM and values used in the site's emergency dose assessment methodology.
* Depending upon the methodolog)'
used to calculate the EAL values , there may be overlap of some values eetween different ICs. Developers will need to address this overlap ey adjusting these values in a manner that ensures a logical escalation in the EGL. The " site specific dose receptor point" is the distanee(s) a:Rd/or locations used ey the licensee to distinguish eet>.veen on site and offsite doses. The selected distanee(s) and/or locations should reflect the content of the emergency plan , and procedural methodology used to determine offsite doses and Protective Action Recommendations. The 1 1ariation in selected dose receptor points means there ma)' ee some differences in the distance from the release point to the calculated dose point from site to site. Developers should research radiation monitor design documents or other inforn,atioA sources to ensure that l) the Ei\L 1 ,ralue eeing considered is within the usable response and display raAge of the instrument , and 2) there are no automatic features that may render the monitor reading iAvalid (e.g., an auto purge feature triggered at a particular indication level). It is recognized that the condition deserieed ey this IC may result in a radiological effluent value eeyond the operatiAg or display range of the installed effluent monitor. In those eases , EAL values should ee determined with a margin sufficient to ensure that an accurate monitor reading is availaele.
For eJrnmple, an EAL monitor readiAg might ee set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding, if the estimated/calculated monitor reading is greater than apprmdmately 110% of the highest accurate monitor reading , then developers may choose not to iAelude the monitor as an indication a:Rd identify an alternate EAL threshold.
Although the IC references TEDE, field suP1ey results are geAerally available onl)' as a "*whole body" dose rate. For this reasoA , the field surve)' EAL specifies a "closed wiAdow" survey reading. IndicatioAs from a real time dose pre:jectioA system are not included in the geAerie EALs. Many licensees do not have this capability.
For those that do, the capability may not be within 66 1>!6 1 99 o J (Revision
: 6) }IO'rember 2 012 the scope of the plant Technical Spec;fi~at1on~:
.~t .. , ill be considered on a case by case basts. real time ~050 prajeetion system res* ts , *rrr ' . . ll , l,s . . *n s *stem are not included m the genenc n . . " licensee may request to include an EAL _using Indications from a penmeter m?~1ton g ) e that do these monitors may not be ~=~:::.i"":::'~:i:i:::.':':.t~:
::::~~~' :::~:'.'
::~~=i::::::~i"!,~::;:nt Technical Specifications.
In addit1?n , ,re;dmg~ L) using a perimeter monitoring system; A * , requestto me u e an ' ' factors. n lwensee m.7 d n a ease by ease basis. pro"al "'ill be eons, ere o ap EGL Assignment Attributes:
3.1.4.C 67 NE I 99 0 1 (R eYisie A 6) J>Je v e FA BeF 2 0 12 AG2RG2 [Sec Dcvc!epa ,\l et-c s] ECL: General Emerge ncy Initiating Condition:
Spent fuel pool level cannot be restored to at least 40 ft.~ in. (site s pecific Level 3 de s cripti o n) for 60_-minutes or longer. Operating Mode Applicability:
All Example Emergency Action Levels: Note: The Emergency Director should declare the General Eme rgency promptly upon determining that 60 minutes ha s been exceeded , or will likely be exceeded.
R Q 2.1 Spent fuel pool level cannot be restored to at least-1.Q ft._]_ in. (s ite specific Level 3 value) for 60 minutes or longer. Definitions:
Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.
It is recognized that this IC would likel y not be met until well after another General Emergency IC was met; however , it is included to provide classification diversity.
De&#xa5;eleper Netes: In accordance with the discussion in Section 1.4 , }JRC Order EA 12 051 , it is recornrnended that this IC and EAL be impl e m e nted wh e n th e enhanced spent fuel pool lev e l instrurnentation i s available for use. The "s ite specific Level 3 value" is usually that spent fuel pool level where fue l remains covered and actions to implement make up 'Nater addition should no longer be deferred.
This site specific level is determined in accordance with NRG Order EA 12 051 and }H~,i 12 02 , and applicable ovmer's group guidance.
Developers should rnodif')' the E AL and/or Ba s is se c tion t o reflect any s ite specific con s traint s or lirnitations as s ociated with the design or o peration of in s trurnentation u s ed to deterrnine the Level 3 value. EGL As s ignment Attribute s: 3.1.4.C 68 1>a;1 99 O I (Revi s ioR 6) Novemser 2012 7 COLD SHUTDOWN/
REFUELING SYSTEM MALFUNCTION ICS/EALS Tobie C 1: Reeegnitien Category "C" Initiating Condition Matrix UNU8U,A .. L EVENT CUl UNPLAN1'rnD loss of (reactor vessel/RCS
[PWR] or RPV [BWR]) inventory for 15 minutes or longer. Op. },fedes: Geld Shuldewn , Refaeli,~gJ:.,_
fr CU2 Loss of all but one AC power source to emergency bHses for 15 minutes or longer. Op. },fedes: LJCeld Shuffiewn , Re.f1,wling, De.fuel-et/
CU3 Ul'~PLt\"NNED increase in RCS temperatllre.
Op. },fetles:
LJCel-tl Shtt:tlewn , Refueling CU4 Loss of Vital DC povter for 15 minutes or longer. Op. },fedes: LJCeld Shuffiewn, Reft1el ing CU5 Loss of all onsite or offsite commHnications capabilities.
Op. Medes: LJCeld Shutdovm , Refaeling , Defaektl ALERT CAl Loss of (reactor vessel/RCS
[0 WR] or RP'' 1 l'--1 [BWR]) inventory.
Op. },fetles:
LJCel-tl Shultlovm , Refueling CAl Loss of all offsite and all onsite AC povt'er to emergency bHses for 15 minutes or longer. Op. 1',1edes:
LJCel-tl Sht1lde>it 1 11 , Refaelin.g , Defael-ed CM Inability to maintain the plant in cold shHtdown.
Op. },fetles:
LJCeld Sht1ldewn , Refueling 69 8ITEAREA EMERGENCY C81 Loss of (reactor vessel/RCS
[PWR] or RPV [BWR]) inventory affecting core decay heat removal capability. Op. 1',fedes:
LJCeld Shutdewn , Refaeling GENERAL EMERGENCY CGl Loss of (reactor vessel/RCS [PWR] or RiDV [BWR]) inventory affecting fuel clad integrity with containment challenged.
Op. },fetle s: LJCeld Sht11tlewn., Refueling 1 Table intended for use by 1 EAL develo13ers.
: Inclusion in licensee I d * . d , ocuments 1s not require . L------------------J UNUSUAL EVENT ALERT CAfi Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Op. },fedes: &#xa3;....fr.Geld Shut-dewn, R(}fucling 70 Nel 99 0 I (ReYisieA
: 6) ~leveA1eer 2012 SITE ARE,A ... GENERl ... L EMERGENCY EMERGENCY I Table intended for use by 1 EAL de 1 relopers.
: Inclusion in licensee : documents is not required.
L------------------J Nel 99 01 (RevisieR
: 6) Nevernber 2012 CU1 ECL: }fotification of Unusual EventUnusual Event Initiating Condition:
UNPLANNED loss of f reactor vessel/RCS
[PWR] or RPV [BWR]) inventory for 15 minutes or longer. Operating Mode Applicability:
Cold Shutdown, Refueling5, 6 Emergency Action Levels: Example Emergeney Aetiee Levels: ( 1 or 2) Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
T l +2 UNPLANNED loss of reactor coolant results ftn f reactor vessel/RCS
[PWR] or R.0 V [BWR]) level less than a required lower limit for 15 minutes or longer. a. f Reactor vessel/RCS
[PWR] or R.nv [BWR]) level cannot be monitored.
AND b. UNPLANNED increase in (site specific sump and/or tank)Containment Sump A OR Waste Holdup Tank levels. Definitions:
UNPLANNED:
A parameter change or an event that is not ]) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inability to restore and maintain water level to a required minimum level ( or the lower limit of a level band), or a loss of the ability to monitor f reactor vessel/RCS
[PWR] or RPV [BWR]) level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that decrease RCS water inventory are carefully planned and controlled.
An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. EAL CUI.I recognizes that the minimum required f reactor vessel/RCS
[PWR] or RPV [BWR]) level can change several times during the course of a refueling outage as diffe r ent plant configurations and system lineups are implemented.
This EAL is met if the minimum level, specified for the current plant conditions , cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.
71 
};El 99 01 (Re,*isieA 6) }!evember 3012 The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. 72 J>Jel 99 0 I (ReYisieB
: 6) J>levemser 2012 EAL CU 1.2 addresses a condition where all means to determine f reactor vessel/RCS
[PWR] or RPV [BWR]) level have been lost. In this condition , operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the f reactor vessel/RCS
[PWR] or RPV [BWR]). Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA 1 or CA3. Develaf)er Nates: EAL #1 It is recogAi2ed that the miAirnum allowable reactor vessel/RCS/RPV level may have maAy values over the course ofa refueliAg outage. Developers should solicit iAput from liceAsed operators concemiAg the optimum wordiAg for this EAL statemeAt.
IA particular, determrn.e if the geAeric v,rordiAg is adequate to eAsure accurate aAd timely classificatioA, or if specific setpoiAts caA be iAcluded without makiAg the EAL statemeAt uAwieldy or poteAtially iACOAsisteAt with aetioAs that may be takeA duriAg aA outage. If specific setpoiAts are iAcluded, these should be drawn from applicaale operatiAg procedures or other coAtrolliAg documeAts.
EAL #2.b EAter aAy "site specific sump aAd/or taAk" levels that could be eM,pected to iAcrease if there \Vere a loss of iAveAtory (i.e., the lost iAventory would eAter the listed sump or taAk). EGL AssigAmeAt Attributes:
3 .1.1.A 73 ECL: Notification of Unusual EventUnusual Event t>I E I 99 01 (RevisieR
: 6) t>l eYeffiber 2012 CU2 Initiating Condition:
Loss of all but one AC power source to emergency buses for 15 minutes or longer. Operating Mode Applicability:
Cold Shutdo*wn , Refueling5, 6 , Defueled Example Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded , or will likely be exceeded.
c&#xb5;2.1 a. AC power capabi li ty to (site specific emergency buses)1(2})-A-05 and 1(2)-A-06 is reduced to a sing l e power source for 15 minutes or longer. AND b. Any additional single power source fai lur e will result in loss of all AC power to SAFETY SYSTEMS. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. Basis: Note: with respect to this EAL. " Station Blackout is Unit 1(2) specific." T his IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition , the sole AC power source may be powering one, or more than one , train of safetyrelated equipment.
* Normal Unit 1(2) offsite power sources include: o 345 KVAC I(2)X-03 through the 13.8 KVAC system to the LVSAT 1(2)X-04 o 345 KVAC backfed through the 19 KVAC system to the UAT 1(2)X-02
* Normal Unit 1(2) onsite power sources consist of: o emergency diesel generators o gas turbine generator o unit main turbine generator o power supplied from the opposite unit When in the cold shutdown, refueling, or defueled mode , this condition is not classified as an A lert because of the increased time available to restore another power source to service. 74 l>JEI 99 01 (ReYisieA
: 6) l>Jevemller
?Q 12 Additional time is available due to the reduced core decay heat load , and the lower temperatures and pressures in various plant systems. Thus, when in these modes , thi s condition is considered to be a potential degradation of the level of safety of the plant. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. 75 
}lei 99 0 I (RevisieA
: 6) }Jeveffieer 2012 An "AC power source" is a source recognized in AOPs a nd EOPs (including Beyond Design Basis event procedures), and capab l e of supp l ying requ ir ed power to an emergency bus. Some exa mpl es of this Initiating condition Condition are presented below.
* A loss of a ll offsite power with a concurrent failure of all but one emergency power source (e.g., an o n site diesel generator).
* A loss of all offsite power and loss of a ll emergency power sources (e.g., onsite diesel generators) with a single train of emergenc y buses being back-fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a s in gle train of emergency buses being back-fed from a n offs i te power source. The s ub sequent loss of the remaining sing l e power source wou ld escalate the event to an A lert in accordance w ith IC CA2. De..*elepeF Netes: for a povi1er source that has multiple generators , the EAL and/or Basis section should reflect the minimum numeer of operating generators neeessar;*
for that source to proYide required power to an AC emergene;r eus. for eM.ample, if a eaekup power source is comprised of tv*,o generators (i.e., hNo 50% capacity generators sized to feed 1 AC emergency eus), the EAL and Basis section must speeif;* that eoth generators for that souree are operating.
The "s ite specific emergenc;r euses" are the euses fed e;r offsite or emergene;r AC power sources that supply power to the electrical distrieution system that pov1ers SAfETY SYSTEMS. There is ty13icall;r 1 emergenc;*
eus per train of SAfETY SYSTEMS. Developers should modif;* the eulleted e>rnmples pro 1 1ided in the easis section, above, as needed to reflect their site specific plant designs and eapaeilities.
The EALs and Basis should reflect that each independent offsite power circuit constitutes a single power source. For e>rnmple, three independent 345kV offsite power circuits (i.e., incoming power lines) comprise three separate power sources. Independence may ee determined from a review of the site specific UFS.AR , SBO analysis or related loss of electrical pov,*er studies. The EAL and/or Basis section ma;r specify use of a non safety related pmver source provided that operation of this source is recognized in AOPs and EOPS, or eeyond design easis accident response guidelines (e.g., FLEX support guidelines).
Such power sources should generall;r meet the "Alternate ac source" definition provided in 10 CFR 50.2. At multi unit stations, the EALs may credit compensatory measures that are proceduralized and can ee implemented within 15 minutes. Consider capaeilities such as pO\Yer source cross ties, "sv,*ing" generators, other power sources desorieed in aenormal or emergeno;*
operating procedures , etc. Plants that ha 1 t1e a proceduralized capaeilit;r to supply offsite AC power to an affected unit via a cross tie to a companion unit may credit this power source in the EAL proYided that the planned cross tie strateg;r meets the requirements of 10 CFR 50.6 3. EGL Assignment Attrieutes:
3 .1.1.A 76 ECL: Notification of Unusual EventUnusual Event Initiating Condition:
UNPLANNED increase in RCS temperature.
Operating Mode Applicability:
Cold Shutdown, Refueling5.
6 Emergency Action Levels: 1-IEI 99 0 I (RevisioA
: 6) 1-lovem l3er 2012 CUJ Example Emergeeey Aetioe Levels: (1 or 2) Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded , or will likely be exceeded.
~1 ~2 UNPLANNED increase in RCS temperature to greater than (site specific Technical Specification cold shutdown temperature lirnit)200&deg;F. Loss of ALL RCS temperature and f reactor vessel/RCS
[PWR] or RPV [BWR]) level indication for 15 minutes or longer. Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment and its associated structures.
systems. and components as a functional barrier to fission product release under shutdown conditions.
Basis: This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and l evel, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director sho uld a l so refer to IC CA3. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function i s ava ilabl e does not warrant a clas s ification.
EAL CU3.1 involves a lo ss of decay heat removal capabi lity , or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the co ld shutdown temperature limit specified in Technical Specifications.
During this condition, there is no immediate threat of fuel damage because the core decay heat load ha s been reduced s inc e the cessation of power operation.
77 l>IEI 99 0 I (Re\'isieA
: 6) Nevember 2012 During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled.
A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.
78 l>lel 99 0 I (Revisien
: 6) l>/evember 2012 EAL CU3.2 reflects a condition where there has been a s i gnificant lo ss of instrumentation capability necessary to monitor RCS conditions and operators wou ld be unable to monitor key parameters necessary to assure core decay heat removal. During this condition , there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Esca lat ion to Alert wo uld be via IC CAI based on an in ventory l oss or IC CA3 based on exceeding plant configuration-specific time criteria.
Developer Notes: For EAL #1, enter the " site specific Technical Specification cold shutdovm temperature lirnit" where indicated. EGL Assignment Attributes:
3 .1.1.i\ 79 ECL: Notification of Unu s ual EventUnu s ual E v e nt Initiating Condition:
Loss of Vital DC power for 15 minutes or longer. Operating Mode Applicability:
Cold Shutdovm , Refuelingl:__Q Examf)le EmergeeeyEmergency Action Levels: J>Je l 99 0 I (R e~*i s ie A 6) J>J eveA'lber 2 0 1 2 CU4 Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded , or will likely be exceeded. C 4.1 Indicated voltage is less than (s ite s pe c ifi c bu s volta g e 11alue) 115 VDC on required Vital DC buses D-0 I. D-0 2 , D-03. or D-04 for 15 minutes or longer. Definitions:
SAFETY SYST E M: A s ys t e m required for s afe plant operation. cooling down the plant and/o r placing it in th e cold shutdo w n conditi o n , including the E CCS. Thes e sys t e ms are c lassifi ed a s safety-related. Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus , this condition is considered to be a potential degradation of the level of safety of the plant. As used in t hi s EAL, " required" means the Vital DC buses necessary to support operation of the in-service , or operable, train or tra in s of SAFETY SYSTEM equ ipm ent. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification. T he safety-rel a ted 125 VDC sy stem con s i s t of four main bu ses: D-0 L D-02. D-03. and D-04. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CAI or CA3 , or an IC in Recognition Category AR. Develaf)er Nates: The " site specific bus voltage value" sh o uld be ba s ed on the miniRrnm bus 11oltage necessary for adequate operation of SAF E TY SYST E M equipment.
This 11oltage 11alue should incorporate a margin of at least 15 minute s of operation before the on s et of inability to operate 80 3.1.1.A 81 l>J E I 99 01 (ReYi s ieA 6) NeYefl'!ser 2012 ECL: Notification of Unusual EventUnusual Event Initiating Condition:
Loss of all onsite or offsite communications capabi liti es. Operating Mode Applicability:
Cold 8hutdov.n , Refueling5, 6 , Defueled Emergency Action Levels: A,etiee LeYels: (1 or 2 or 3) C 5.1 Loss of ALL of the following onsite communication methods: CU5 _* _(site specific list of communications methods)Plant Public Address System Tronics) *
* Commercial Phones
* PBX Phones
* Security Radio
* Portable Radios C 5.2 Loss of ALL of the following GRGoffsite response organization communications methods:
* Nuclear Accident Reporting System (NARS)
* Commercial Phones
* PBX Phones
* Satellite Phones
* Manitowoc County Sheriffs Department Radio C 5.3 Loss of ALL of the following NRC communications methods: _* _(site specific list of communications methods)FTS Phone System
* Commercial Phones System
* PBX Phones
* Satellite Phones Definitions:
None Basis: Th i s IC addresses a significant loss of on-site or offsite communications capabilities.
While not a direct cha ll enge to plant or personne l safety, this event warrants prompt notifications to GRQoffsite response organization s and the NRC. Th i s IC s hould be assessed only w h en extraordinary means are being utilized to make communications possible (e.g., use of non-plant , privately owned equipment, relaying of on-site information via individua l s or multiple radio transmission points , indi viduals being sent to offsite l ocations, etc.). 82 
+>le i 99 O I (RevisieA
: 6) t>l eveffieer 2012 EAL CU5.l addresses a total loss of the communications methods used in support of routine plant operations.
EAL CU5.2 addresses a total loss of the communications methods used to notify all GRGoffsite response organ i zation s of an emergency declaration. The offsite response organizations referred to h ere are the State of W i sconsin, Manitowoc County. and Kewa u nee CountyThe OROs referred to here are (see Developer Notes). ---EAL CU5.3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. DevelefJeF Netes: EAL #1 The " site specific list of communications methods" should include all communications methods used for routine plant communications (e.g., commercial or site telephones , page part)' systems , radios , etc.). This listing should include installed plant equipment and components , and not items o*wned and maintained by individua l s. EAL #2 The " site specific list of communications methods" should include all communications methods used to perform initial emergency' notifications to OROs as described i n the s i te Emergency Plan. The listing should include installed plant equipment and components , and not items O'+'med and maintained by individuals.
EM.ample methods are r i ng down,ldedicated telephone lines, commercial te l ephone lines , radios , satellite telephones and i nternet based communications techno l ogy. 1A the Basis section , insert the site specifi c listing of the OROs requiring notification of an emergency' declaration from the Control Room in accordance with the site Emergency Plan , and typically within 15 minutes. EAL #3 The " site specific list of communications methods" should include all communications met h ods used to perform initia l emergency notifications to the *NRG as described in the site Emergency Plan. The l isting sho uld include i nstalled plant equipment and components , and not items ovmed and maintained by individuals.
These methods are t)'pieally the dedicated Emergency Notification System (E~JS) telephone line and commercial telephone lines. EGL Assignment Attributes:
3.1.1.C 83 
~JEI 99 01 (Revi s i e R 6) Neveffiber
?Q 12 CA1 ECL: Alert Initiating Condition:
Loss of f reactor vessel/RCS
[PWR] or R.0 V [BWR]) inventory. Operating Mode Applicability:
Cold Sh u tdown , Refueling5, 6 Emergency Action Levels: ExamJJle EmergeueyEmergeuey f..etien Levels: (1 or 2) Note: The Emergency Director should declare the Alert promptly upon determining that 15_ minutes has been exceeded, or will likely be exceeded.
~1 ~2 Loss of f reactor vessel/RCS
[PWR] or R.0 V [BWR]) inventory as indicated by level less than (site specific level)l 6% on Ll-44 7 / Ll-447A. a. f Reactor vessel/RCS
[PWR] or R.0 V [BWR]) level cannot be monitored for 15 minutes or longer b. AND UNPLANNED increase in (site specific sump and/or tank)Containment Sump A OR Waste Holdup Tank levels due to a loss of f reactor vessel/RCS
[PWR] or RPV [BWR]) inventory.
Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses cond i tions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precur sor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety. For EAL CA 1.1 , a lowering of water level below (site specific leve l) 16% on LI-447 I LT-447 A indicates that operator actions have not been successful in restoring and maintaining f reactor vessel/RCS
[PWR] or RPV [BWR]) water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery. The LI-447/LI-447A threshold corresponds to the min i mum shutdown reactor vesse l l eve l required for operation of RHR without a i r binding the suction. Although related , EAL CA I .1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal ( e.g., loss of a Residual Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. 84 
}le i 99 0 I (R ev i s i o n 6) }le v e me e r 2012 For EAL #1 the LI 447/LI 447A threshold corresponds to 6 inches above the bottom ins i de diameter of the RCS loop. Thi s condition will result in a minimum classification of Alert. For EAL CAI .2, the inability to monitor f reactor vesse l/R CS [PWR] or RPV [BW1"]) level may be ca u sed by in strumentation and/or power fa ilur es , or water level dropping below the range of ava il able instrumentation.
If water l evel cannot be monitored , operators may determine t hat an inventory loss i s occurring by observing changes in sump and/or tank levels. Sump and/or tank l eve l changes must be eva lu ated against other potential sources of water flow t o ensure they are indicative of l eakage from the f reactor vesse l/RC S [PWR] or RPV [B~&#xa5;R]). 85 l>Je l 99 O 1 (RevisieA
: 6) 1>J e\'effi'9er 2012 The 15-minute duration for the lo ss of level indi cation was chosen becau se it is half of the EAL duration specified in IC CS 1 If the f reactor vessel/RCS
[PWR] or RPV [BWR]) inventory level continues to l ower, then escalation to Site Area Emergency would be via IC CS 1. DeYeleper Netes: For EAL #1 the " site specific level" should be based on either: * [BWR] Low Lov,r EGGS actuation setpoint/Level
: 2. This setpoint *.vas chosen because it is a standlli'd operationally significant setpoint at which some (typically high pressure EGGS) injection systems v,ould automatically start and is a value sigAificantl;
' below the low R..0 V water level RPS actt:1ation setpoint specified in IC CU 1. * [PWR] The minimum allowable level that supports operation ofnormall;'
used decay heat removal systems (e.g., Residual Heat Removal or Shutdovm Cooling).
If multiple levels e>dst, specify each along *uith the appropriate mode or configuration dependency criteria. For EAL #2 The type and range of RCS level instrumentation may vary during an outage as the plant moves throttgh variotts operating modes and refueling evoltttions , plli'ticularly for a PWR. As appropriate to the plant design , alternate means of determining RCS level are installed to assure that the ability to monitor le,,el within the range reqttired by operating procedures will not be interrupted.
The instrumentation range necessary to sttpport implementation of operating procedures in the Cold Shutdown and Refueling modes may be different (e.g., narrower) than that reqttired during modes higher than Cold Shutdown.
Enter any " site specific sump and/or tank" le*,els that could be e>{pected to increase ifthere were a loss of inventory (i.e., the lost inventory would enter the listed sttmp or tank). EGL AssigAment Attribtttes:
3 .1.2.B 86 ECL: Alert N E I 99 0 I (Re\*i s i o A 6) 1>Jo YeFAber 2012 CA2 Initiating Condition:
Loss of all offsite and all onsite AC pow e r to emergency bu ses for 15 minutes or longer. Operating Mode Applicability:
Cold Shutdovm , Refueli11g5.
6 , Defueled Emergency Action Levels: Example EmergeneyEmergeney Aetion LeYels: Note: The Emergency Director should declare the Alert promptly upon determining that 15_ minutes has been exceeded, or will likely be exceeded. C 2.1 Loss of ALL offsite and ALL onsite AC Power to (site specific emerge11cy buses)ill)::
A-05 and ](2)-A-06 for 15 minutes or lon ge r. Definitions:
SAFETY SYSTEM: A system required for safe plant operation.
cooling down the plant and/or placing it in the cold shutdown condition. including the ECCS. These systems are classified as safety-related.
Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessar y for emergency core cooli n g, containment heat removal/pressure control , spent fuel heat removal and the ultimate heat sink. When in the cold shutdown, refueling , or defueled mode , this condition is not cla ss ified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load , and the lower temperatures and pressures in various plant systems. Thus, when in these modes , this condition represents an actual or potential substantial degradation of the level of safety of the plant. For the purpose of classification under this EAL, evaluation of power sources should be made on each unit individually.
---If mitigative strategies establish emergency power to any bus listed in the EAL, the EAL threshold for this Initiating Condition is not met. Fifteen minutes was selected as a thre s hold to exclude transient or momentary po we r losses. ---Esca lation of the emergency cla ss ification l eve l would be via IC CS 1 or AS+RS 1. 87 Develeper Netes: }/E l 99 0 I (Rev i sieA 6) }l evember 2012 For a power source that has multiple g eAerators , the EAL aAd/or Basis sectioA should reflect the miAimum number of operatiAg generators Aecessary for that source to provide adequate power to aA AC emergency bus. For e1rnmple, if a backup power source is comprised of two generators (i.e., t\Yo 50% capacity geAerators sized to feed 1 AC emergency bus), the EAL and Basis sectioA must specify that both geAerator s for that source are operatiAg.
The " site specific emergeAC)' buses" are the buses fed by offsite or emergeAC)' AC po 1 ,&#xa5;er sources that supply power to the electrical distributioA system that powers SAFETY SYSTEMS. There is typically 1 emergency bus per train of SAFETY SYSTEMS. The EAL and/or Basis section may specif)' use of a non safety related po',Yer source pro 1 1ided that operation of this source i s controlled in accordance with abnormal or emergency operating procedures , or beyond desigA basis accident response guidelines (e.g., FL EX support guidelines). Such power sources should geAerally meet the " Alternate ac source" defiAitioA provided iA 10 CFR 50.2. At multi unit statioAs , the E,r\Ls may credit compensatory measures that are proceduralized and can be implemented within 15 minutes. Consider capabilities such as power source cross ties, " swing" generators , other power sources described in abnormal or emergency operating procedures , etc. Plants that have a proceduralized capability to supply offsite AC power to an affected Hnit Yia a cross tie to a companion unit may credit this power source in the E AL provided that the planned cross tie s trategy meets the requirement s of 10 CFR 50.63. EGL Assignment AttribHtes:
3.1.2.B 88 1'16 1 99 0 1 (R e vi s i e A 6) 1'1e ,'e ffiber 2 0 12 CA3 ECL: Alert Initiating Condition:
Inability to maintain the plant in cold shutdown.
Operating Mode Applicability:
Cold Shutdown , Refueling5.
6 Emergency Action Levels: Example Emergeeey A.etioH Levels: (1 or 2) Note: The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded , or will likely be exceeded. ~l C 3.2 UNPLANNED increase in RCS temperature to greater than (site specific Technical Specification cold shutdown temperature limit)200&deg;F for greater than the duration specified in the following table. Table: RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Heat-up Duration Status Intact (but not at reduced Not applicable 60 minutes* inventory
[PWR]) Not intact (or at reduced Established 20 minutes* inventory fPWR]) Not Established 0 minutes
* _If an RCS heat remoYal systemRHR is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
UNPLANNED RCS pressure increase greater than (site specific pressure reading)25 l!fil_g. (This EAL does not apply during water-solid plant conditions.
[PWR]) Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment and its associated structures.
systems. and components as a functional barrier to fission product release under shutdown conditions.
Basis: This IC addresses conditions involving a loss of decay heat removal capability or an add i tion of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant. 89 N e l 99 01 (ReYi s i e A 6) ~l 0~'0ffi00F 2012 A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limi t when the heat removal function is avai l able does not warrant a classification.
90 l>le l 99 0 I (Re~*i s i e R 6) l>Je v e ffie e r 2 0 12 The RCS Heat-up Duration Thresho ld s table addresses an in crease in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact , or RCS inventory is reduced (e.g., mid-loop operation in PWRs). The 20-minute criterion was included to a ll ow time for operator action to address the temperature increase.
The RCS Heat-up Duration Thresho ld s table a l so addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety. Fina ll y , in the case where there is an increase in RCS temperature , the RCS i s not intact or is at reduced inventory
[PWR], and CONTAINMENT CLOSURE is not established , no heat-up duration is al l owed (i.e., 0 minutes).
This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequent l y to the environment , and 2) there is reduced reactor coo l ant inventory above the top of irradiated fuel. EAL CA3.2 provides a pressure-based indication of RCS heat-up. Esca lat ion of the emergency classification level would be via IC CS 1 or A&l-RS I. DeYelef)er Netes: For EAL #1 Enter the "s ite specific Technical Specification cold shutdovm temperature limit" where indicated.
The RCS should be con s idered iAtact or not intact in accordaAce with site specific criteria.
For EAL #2 The "s ite specific pressure reading" should be the lov, , est change in pre s sure that caA be accurately determined u s ing installed iAstrumentatioA , but not less than 10 psig. For PWRs , this IC and its a ss ociated EA.Ls address the concerns raised by Generi c Letter 88 17 , Less efDee*ly' Heet R e me*;el. A number of phenomena such as pressurization , vorte1,ing , steam generator U tube drainiA g, RCS level differences when operating at a mid loop condition , decay heat removal system design , and level instrumentation problems can lead to conditions where decay heat removal is lost and core unc o very can oc c ur. *NRG aAalyse s s ho*,v that there a r e sequences that can cau s e core uAcovery iA 15 to 20 minutes , aAd severe core damage withiA an hour after decay heat removal is lost. The allowed time frames are coA s istent with the guidaAce provided by Generic Letter 88 17 and believed to be coA s ervative giveA that a low pressure CoAtaiAment barrier t o fi ss ioA product release i s establi s hed. EGL Assignment Attributes:
3.1.2.B 91 
}IE! 99 01 (RevisieA
: 6) }l eveA'l!ieF 2012 CA6 ECL: Alert Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability:
Cold Shutdovm , Refueling5, 6 Emergency Action Levels: Example Emergeney A,etian LeYels: ,;_ Notes: C jt\6.1
* If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
* If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then thi s emergency classification is not warranted. a. The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* (site s13ecific hazards)Lake level greater than or equal to 9.0 ft. (Plant elevation)
* Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director AND b. EITHER of the following:
------1. E vent damage has caused indications of degraded performance in at least one train of a SAFETY SYST E M needed for the current operating mode. 2. 2EITHER of the following:-:-* Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, or 92 
~IEI 99 0 I (Re~*isieA 6) Nevemser 2012
* The event has caused resulted in VISIBLE DAMAGE to the second train of aa SAFETY SYSTEM compoAeAt or structure needed for the current operating mode. 93 Definitions:
Wei 99 0 I (Re\*i s ion 6) ~lo,*ember 2012 FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
EXPLOSION:
A rapid. violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction.
or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits.
grounding, arcing, etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. SAFETY SYSTEM: A system required for safe plant operation.
cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Damage resulting from an equipment failure and limited to the failed component (i.e .. the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train. and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words. in order for this EAL to be classified.
the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance.
and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria CA6.1.b.1 of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or re l iability of the SAFETY SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. 94 1>161 99 0 I (RevisieA
: 6) l>leveffiber 2012 This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM , or a structure containing SAFETY SYSTEM components , needed for the current operating mode. This condition significantly reduces the margin to a lo ss or potential loss of a fission product barrier , a11d therefore represents an actual or potential substantial degradation of the level of safety of the EAL l.b.l addresses damage to a SA::.4 ETY SYSTEM train that is in service/operation since indications for it will be read ii:,* available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. EAL l .b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SA.FETY SYSTEM components.
Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC CS l or ASl RS 1. Devel0peF N0tes: For (site specific hazards), developers should consider includin g other significant, site specific hazards to the bulleted list contained in EAL 1.a (e.g., a seiche). *Nuclear power plant SAFETY SYSTEMS are comprised of two or more separate and redundant trains of equipment in accordance with site specific design criteria.
EGL /\ssignment Attributes:
3.1.2.B 95 
}l e i 99 01 (Revi s i e A 6) }l e*,*effieer 2012 CS1 ECL: Site Area Emergency Initiating Condition: Loss of f reactor vessel/RCS
[PWR] or R..0 V [BTf!R]) inventory affecting core decay heat removal capability.
Operating Mode Applicability:
Cold Shutdown , Refueling5.
6 Emergency Action Levels: Example Emergeney Aetion Levels: (1 or 2 or 3) Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
: b. (Reactor vessel/RCS
[PWR] or R..0 V [BWR]) )eye) less than (site specific level) b. (Reactor Yessel/RCS
[PWR] or R.0 V [BWR]) level less than (site specific level). ~fil 31 a. f Reactor vessel/RCS
[PWR] or R..0 V [BWR]) level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is i nd i cated by ANY of the following: Definitions:
* (Site specific radiation monitor)Containment High Radiation Monitor. -fl..(21=
RE-126. 1(2)RE-127.
or 1(2)-RE-128.;t reading greater than (site specific ~lOOR/hr
* Erratic source range monitor indication
[PWR]
* UNPLANNED increase in (site specific sump and/or tank)Containment Sump A OR Waste Holdup Tank l evels of sufficient magnitude to indicate core uncovery * (Other site specific indications)
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. 96 Basis: }l e i 99 01 (Re\'ision 6) }Jovemeer 2012 This IC addresses a sig nificant and prolonged lo ss of f reactor vessel/RCS
[PWR] or RPV [BWR]) inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventor y may be due to a RCS component failure, a lo ss of configuration control or prolonged boilin g ofreac tor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Eme r gency d eclaration. Following an extended los s of core d ecay heat rem ova l and inventory makeup , dec ay h ea t will cause reactor coolant boiling and a further reduction in reactor vessel l evel. If R CS/reactor vessel level cannot be restored , fuel damage is probable.
Outage/slmtdovm contingency plans typically provide for re establishing or verifying CONTAil'lMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs 1.b and 2.b reflect the fact that *with CONTAINMENT CLOSURE established , there is a lower probability of a fission product release to the environment.
.. In EAL CS l .JJ.a , the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows s ufficient time to monitor , assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainti es). It also allows sufficient time for performance of actions to terminate leaka ge, recover inventory control/makeup equipment and/or restore level monitoring. For EAL CS 1.-; I.a, the calculated radiation l evel on the Containment High Radiation Monitors (RE-126, RE-127. or RE-128) is without the reactor head in place. Calculated radiation l evels with the reactor head in place are below the usable scale of these monitors.
The inability to monitor f rea ctor vessel/RCS
[PWR] or RPV [BWR]) level may be ca use d b y instrumentation and/or power failures , equipment not calibrated for the plant conditions (e.g., hot cal only). or water level dropping below the range of available instrumentation.
If water level cannot be monitored, operators may determine that an inventor y loss i s occ urr i ng by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ens ure th ey are indicative of leak age from the f reactor vessel/RCS
[PWR] or R.0 V [BWR]). These EALs address concerns raised b y Generic Letter 88-17 , Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Ri sk Issues; NUREG-1449 , Shutdown and Low-Power Op e ration at Commercial N ucl ear Power Plants in the United States; and NUMARC 91-06 , Guide lin es for Indu stry Actions to Assess Shutdown Managem e nt. ---Esca lation of the emergency classification l eve l would be via IC CG 1 or AG+RG I. Develeper Netes: Accident analyses suggest that fuel damage may occur within one hour of uncovery depending upon the amount of time since shutdown; refer to Generic Letter 88 17 , SECY 91 283, NUREG 1449 and NUMARC 91 06. 97 
}JE;I 99 0 I (Re v i s i e A 6) }l evemeer 2012 The type and range of RCS level instrumentation may vary during an outage as the plant moves through various operating modes and refueling evolutions , particularly for a PWR. As appropriate to the plant design , alternate means of determining RCS level are installed to assure that the ability to monitor level within the range required by operating procedures
*.vill not be interrupted.
The instrumentation range necessary to support implementation of operating procedures in the Cold Shutdovm and Refueling modes may be different (e.g., narro 1.ver) than that required during modes higher than Cold Shutdovm.
For EAL #1.b the " site specific level" is 6" below the bottom ID of the RCS loop. This is the level at 6" belov, the bottom ID of the reactor vessel penetration and not the lovi' point of the loop. If the availabilit)
' of on scale level indication is such that this level value can be determined during some shutdown modes or conditions , but not others , then specify the mode dependent and/or configuration states during which the level indication is applicable. If the design and operation of water level instrumentation is such that this level value cannot be determined at any time during Cold Shutdovm or Refueling modes, then do not include EAL #1 (classification
,,,,,ill be accomplished in accordance with EAL #3). For EAL #2.b The " site specific level" should be approximately the top of active fuel. If the availabilit)'
of on scale level indication is such that this level value can be determined duriH.g some shutdov,n modes or conditions, but not others, then specify the mode dependent and/or configuration states during which the leYel indication is applicable.
If the design and operation of v,ater level instrumentation is such that this le*,el value cannot be determined at any time during Cold Shutdo*.vn or Refueling modes, then do not include EAL #2 (classification will be accomplished in accordance
*with EAL #3). For EAL #3.b first bullet As water leYel in the reactor vessel lo,Ners , the dose rate above the core will increase.
Enter a " site specific radiation monitor" that could be used to detect core uncoyery and the associated
" site specific Yalue" indicative of core uncovef)'.
It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases , EAL *,alues should be determined with a margin sufficient to ensure that an accurate monitor reading is available. For ex.ample , an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than apprm!.imately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold.
To further promote accurate classification , developers should consider if some combination of monitors could be specified in the EAL to build in an appropriate level of corroboration betv, , een monitor readings into the classification assessment.
For EAL #3.b second bullet Post TMI accident studies indicated that the installed PWR nuclear instrumentation
*.viii operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
For EAL #3.b third bullet Enter any 'site specific sump and/or tank" levels that could be eJl.pected to change if there were a loss of RCS/reactor vessel inventory of sufficient magnitude to indicate core uncovery.
Specific level values may be included if desired. 98 1>J~I 99 0 I (Re~*isieA 6) }JSY0FA00F 20 ( 2 For BAL #3.b fourth bullet Developers should determine if other reliable indicators e1(ist to identify fuel uncovery (e.g., remote viewing using cam.eras).
The goal is to identify any unique or site specific indications, not already used else*.vhere, that will promote timely and acc u rate emergency classification.
For BAL # 1.b "site specific level" is the Low Lov, 1 Low EGGS actuation setpo i nt / Level 1. The B\VR Low Low Lev, EGGS actuation setpoiAt / Level 1 was choseA because it is a staAdard operationall;y sigAificant setpoiAt at which some (typicall; 1 lov,r pressure EGGS) iAjection systems 1 Nould automatically start aAd attempt to restore RPV level. This is a R.0 V water level value that is observable below the Lov,r Low/Level 2 value specified iA IC CAI, but significaAtly above the Top of Active Fuel (TO,'\c..v) threshold specified in EAL #2. For EAL #2.b The " site specific level" should be for the top of active fuel. For EAL #3.b first bullet As water level in the reactor vessel lowers, the dose rate above the core *Nill iAcrease.
Enter a "site specific radiation monitor" that could be used to detect core uAcovery aAd the associated
" site specific value" iAdicative of core uAcovery.
It is recognized that the conditioA described b;1 this JC may result iA a radiation value beyoAd the operatiAg or display range of the installed radiatioA monitor. In those cases, EAL values should be determined with a margiA sufficient to eAsure that an accurate monitor reading is available. For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding, if the estimated/calculated monitor reading is greater than apprmcimately 110% of the highest accurate moAitor reading, then developers may choose not to iAclude the monitor as an indication aAd identify aA alternate EAL threshold.
To further promote accurate classificatioA , developers should coAsider if some combination of moAitors could be specified in the EAL to build in aA appropriate level of corroboration between monitor readings into the classification assessment.
For BWRs that do not have installed radiation monitors capable of indicating core uncovery, alternate site specific level indications of core uncovery should be used if available. For EAL #3.b second bullet Because BWR source raAge monitor (SRM) nuclear instrume0tation detectors are typically located below core mid pla0e, this may 0ot be a viable indicator of core u0eovery for BWRs. For EAL #3.b third bullet E0ter aAy "site specific sump a0d/or taAk" leve l s that could be e1cpeeted to chaAge if there v,*ere a lqss of RPV inventory of sufficie0t mag0itude to indicate core uncovery. Specific level values may be included if desired. For EAL #3.b fourth bullet Developers should determiAe if other reliable iAdicators e1dst to identify fuel uAcovery (e.g., remote viewing usiAg cameras).
The goal is to ide0tify any unique or site specific indicatioAs , Rot already used else 1 uhere , that will promote timely aAd accurate emerge0cy classificatioA.
EGL Assignme0t
/\ttributes:
3 .1.3 .B 99 
~l e i 99 01 (ReYi s ieA 6) Nevember 2012 CG1 ECL: General Emergency Initiating Condition:
Loss of f reactor vessel/RCS
[PWR] or R..0 V [BW~]) inventory affecting fuel clad integrity with containment challenged.
Operating Mode Applicability:
Cold Shutdovm , RefueliAg5.
6 Emergency Action Levels: Example EmergeneyEmergenev A,etian Lenis: (1 or 2) Note: The Emergency Director should declare the General Emergency promptly upon determining tha t 30 minutes has been exceeded , or will likely be exceeded.
AND } ..... iAdicatioA fro1H the CoH:taiAmeAt ChalleAge Table (see below). C 1.+/-l a. f Reactor vessel/RCS
[PWR] or RPV [BW~]) level cannot be monitored for 30 minutes or longer. ------AND b. Core uncovery is indicated by ANY of the following:
* (Site specific radiatioA moAitor)Containment High Radiation Monitors.
1(2)RE-126.
](2)RE-127. or 1(2)RE-128, (1(2) RE 126, RE 127. or RE 128) reading greater than 100 R/hrCoAtaiAmeAt High RaAge MoAitor reading greater than (site specific value)..Bthr
* Erratic source range monitor indication
[PWR]
* UNPLANNED increase in (site specific sump and/or taAk)Containment Sump A OR Waste Holdup Tank levels of sufficient magnitude to indicate core uncovery * (Other site specific indications)
AND c. ANY indication from the Containment Challenge Table (see belov,r)~. Containment Challenge Table C-1
* CONTAINMENT CLOSURE not established*
* (E 1 (plosive mi1(ture)6%
H 2 exists inside containment
*-UNPLANNED increase in containment pressure
* SecoAdary containmeAt radiation moAitor reading above (site specific value) [BWR] 100 
}IEI 99 01 (RevisieA
: 6) }levemeer 2012
* If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-_minute time limit , then declaration of a General Emergency is not required.
101 Definitions:
1>J e l 99 01 (Revi s ieA 6) 1>1 evefflber 2012 CONTAINMENT CLOSURE: Th e procedurally defined conditions or actions taken to secure containment and i ts associated structures, systems, and components as a functional barrier to fission product re l ease under shutdo w n conditions. UNPLANNED:
A parameter chang e or an event that is not 1) the result o f an intended evolution or 2) an e x pected plant re s pon se to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged.
This condition represents actual or IMMINENT substantial core degradation or me l ting with potential for loss of containment integrity.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup , decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored , fuel damage is probable.
With CONTAINMENT CLOSURE not established , there is a high potential for a direct and unmonitored release of radioactivity to the environment.
If CONTAINMENT CLOSURE is established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
The existence of an explosive mixture means , at a minimum , that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.
It therefore represents a challenge to Containment integrity.
In the early stages of a core uncovery event , it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment.
If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage , it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service , operators may use the other listed indications to assess whether or not containment is challenged.
In EAL CG l .~-1.~, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor , assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time for performance of actions to terminate leakage , recover inventory control/makeup equipment and/or restore level monitoring. For EAL CG 1.i 1.b , the calcu l ated radiation level on the Containment High Radiat i on Monitors.
102 
~J!;I 99 01 (ReYisieA
: 6) ~Jevemser 2012 I(2)RE-l 26, 1(2)RE-l 27. or 1(2)RE-I 28. (1(2) RE 126, RE 127. or RE 128) is without the reactor head in place. Calculated radiation levels with the reactor head in place are below the usable scale of these monitors.
10 3 NE I 99 0 I (R e ,*i s i o A 6) T>lo ,*e mser 2012 The inability to monitor f reactor vessel/RCS
[PWR] or RPV [.BWR]) l evel may be caused by instrumentation and/or power failures. equipment not calibrated for the plant conditions (e.g .. hot cal only), , or water level dropping bel ow the range of ava il able instrumentation.
If water level cannot be monitored, operators may determine that an inventor y loss i s occurring by observing changes in sump and/or tank l evels. Sump and/or tank level changes mu st be evaluated against other potential sources of water flow to ensure they are indicative of leaka ge from the f reactor vessel/RCS
[PWR] or RPV [.BWR]). These EALs address concerns raised by Generic Letter 88-17, Loss of D ec ay Heat R emoval; SECY 91-283 , Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Pow er Operation at Commercial Nuclear Power Plants in th e United States; and NUMARC 91-06 , Guidelines for Indust ry Actions to Assess Shutdown Management.
DevelaJ:Jer Nates: Accident analyses s1:1ggest that fuel damage may occ1:1r within one ho1:1r of 1:1nc0Yery depending 1:1pon the amo1:1nt of time since sh1:1tdovm
; refer to Generic Letter 88 17 , SECY 91 283 , tJUREG 1449 and NUMA.RC 91 06. The type and range of RCS leYel instr1:1mentation may vary d1:1ring an 01:1ta g e as the plant moYes thrn1:1gh vario1:1s operating modes and refoeling evol1:1tions , partic1:1larly for a PWR. As appropriate to the plant design , alternate means of determining RCS leYel are installed to ass1:1Fe that the ability to monitor level \Yithin the range req1:1ired by operating proced1:1res will not be intermpted. The instrumentation range necessary to s1:1pport implementation of operating proced1:1res in the Cold Shutdown and Refueling modes may be different (e.g., narrower) than that req1:1ired d1:1Fing modes higher than Cold Shutdown.
For EAL #1.a The "site specific level" sho1:1ld be apprmdmately the top of actiYe foe!. If the aYailability of on scale le, 1 el indication is s1:1ch that this !eye) val1:1e can be determined during some shutdown modes or conditions , b1:1t not others , then specify the mode dependent andJor configuration states d1:1ring which the level indication is applicable.
If the design and operation of water le1t 1 el instrumentation is s1:1ch that this level Yal1:1e cannot be determined at any time d1:1ring Cold Sh1:1tdown or Refueling modes , then do not include EAL #1 (classification will be accomplished in accordance with EAL #2). For EAL #2.b first bullet As water level in the reactor vessel lowers , the dose rate abo,*e the core 1.&#xa5;ill increase.
Enter a " site specific radiation monitor" that co1:1ld be 1:1sed to detect core 1:1ncovery and the associated
" site specific val1:1e" indicative of core uncover)'. It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases , EAL values should be determined with a margin s1:1fficient to ensl:IFe that an acc1:1Fate monitor reading is available. For ex.ample , an EAL monitor reading might be set at 90% to 95% of the highest acc1:1Fate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than apprmdmately 110% of the highest acc1:1Fate monitor reading , then developers may choose not to incl1:1de the monitor as an indication and identif)1 an alternate EAL threshold. To further promote acc1:1Fate cla s sification , de, 1 elopers should consider if some combination of monitors co1:1ld be specified in the EAL to b1:1ild in an appropriate level of corroboration between monitor readings into the classification assessment.
104 1>11;1 99 0 I (Re~*isieA 6) J>leveffiser 2012 For BWRs that do not have installed radiation monitors capable of indicating core uncovery, alternate site specific level indications of core uncovet)'
should be used if available.
For EAL #2.b second bullet Post TMJ accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
Because BWR Source Range Monitor (SRM.) nuclear instrumentation detectors are typically located belovi' core mid plane, this may not be a viable indicator of core uncovery for B\VRs. For EAL #2.b third bullet Enter any "site specific sump and/or tank" levels that could be expected to change if there v,*ere a loss of inventor)'
of sufficient magnitude to indicate core uncovery.
Specific level values may be included if desired. For EAL #2.b fourth bullet Developers should determine if other reliable indicators e1dst to identify fuel uncovery (e.g., remote vie1,ving using cameras).
The goal is to identify any unique or site specific indications, not already used elsewhere , that will promote timely and accurate emergency classification. For the Containment Challenge Table: Site shutdo 1 ,'lfl contingency plans t)*pically provide for re establishing CO:J)ffAil>JMeNT CLOSURE following a loss of RCS heat removal or inventory control functions.
For "ExplosiYe mb,ture" , developers ma)' enter the minimum containment atmospheric hydrogen concentration necessary to support a hydrogen burn (i.e., the lower deflagration limit). A concurreet containment mfygen concentratioe may be iecluded if the plant has this i0dicatio0 available ie the Coetrol Room. For BWRs, the use of secondary coetaiement radiation monitors should provide indication of increased release that may be indicative of a challenge to secondary containment.
The "site specific value" should be based on the EOP ma1dnmm safe values because these values are easily recognizable and have a defined basis. EGL Assignment Attributes:
3.1.4.B 105 
}l e i 99 O I (Revision
: 6) }lovember 2012 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS Table E 1: Reeegeitiee Categen' "E" IHitiatieg CeHditiee Matrix UNUSUAL EVENT E HUl Damage to a loaded cask CONFINEMENT BOUNDARY. Op. },1edes: All 106 1 -------------------
: Table iHteAEied for use by 1 EAL de,,*elopers. : !AelusioA iA lieeAsee I ...I * * ...I 1 uocumeAts 1s Aot requtreu.
1 L------------------J ISFSI MALFUNCTION ECL: NotificatioA of UAusual EveAtUnusua l Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability:
All Examf)le Emergency Action Levels: E-H U1 E I.I Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than the va l ues shown be l ow(2 t i mes the site specific cask specific techAical specificatioA allowable radiatioA level) on the surface of the spent fuel cask. 32 PT DSC Front Surface 1700 mrem/-hr7 Door Centerline 400 mrem/-hr7 End Shield Wall Exterior 162 mrem/-hr7 VSC-24 Sides 200 mr em/-hr 7 Too 400 m rem/-hr7 Air Inlets 700 mrem/-hr7 Air Outlets 200 mrem/-hr7 Definit i on: CONFINEMENT BOUNDARY:
The barrier(s) between spent fue l and the environment once the spent fuel is processed for dry storage. Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dr y storage beginning at the point that the loaded storage cask i s sealed. The issues of concern are the creation of a potential or actual release path to the environment , degradation of one or more fuel assemblies due to environmental factors , and confi g uration changes which could cause challenges in removing the cask or fuel from storage. The existence of " damage" is determined by radiolog i ca l survey. The techn i cal specification mu l tiple of " 2 times", wh i ch is also used in Recognition Category A-R IC t\-tf+RUl , is used here to distinguish between non-emergency and emergency conditions
._ The emphasis for this classification is the degradation in the level of safet y of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of e x treme damage to a loaded cask , the fact that the " on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. 107 ISFSI MALFUNCTION Security-related events for ISFSis are covered under ICs HUI and HAI. De*1eloper Notes: The results of the ISFSI Safet)* Analysis Report (SAR) [per NUREG 1536], or a SAR refereneed in the eask Certifieate of Complianee and the related l'lRC Safety Evaluation Report , identify the natural phenomena events and aeeident eonditions that eould potentially affeet the CONFINEMEl'ff BOUNDARY.
This EAL addresses damage that eould result from the range of identified natural or man made events (e.g., a drop13ed or tipped over eask , EXPLOSION , FIRE, EARTHQUKE, ete.). The allov,rable radiation level for a spent fuel eask ean be found in the easlc's teehnieal speeifieation loeated in the Certifieate of Complianee. EGL Assignment Attributes:
3.1.1.B 108 9 FISSION PRODUCT BARRIER ICS/EALS Table 9 F 1: Reeognition Category " F" Initiating Condition Matri>{ ALERT Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier. Loss or Potential Loss of any t\vo barriers.
Loss of any two barriers and Loss or Potential Lo ss of the third barrier. See Table 9 F 2 feF BV/R EALs See Table 9 F 3 feF PWR EA.Ls DeYelepeF Nate: The adjaeent logie flow diagram is for use by developers and is not required for site speeifie implementation
; however , a site speeifie seheme must inelude some type of user aid to faeilitate timely and aeeurate elassifieation of fission produet barrier losses and/or potential losses. 8ueh aids are typieally eomprised of logie flow diagrams, " seoring" eriteria or eheekbm&#xa3; type matriees.
The user aid logie must be eonsistent with that of the adjaeent diagram. 109 l>ie l 99 0 I (Re*,i s i e R 6) l>le vemeer 20 I?
0 ...... ...... 
 
I I I I Table 9 F 2: B\\'R EAL Fission Prnduet Barrier Table Thresholds far LOSS or POTENTIAL LOSS of Barriers NEl 99 01 (RevisioR
: 6) November 2012 I FAl ALERT I FSl SITE AREA, EMERGENCY I FG 1 GENERAL EMERGENC" AA)' Loss or Emy PoteAtial Loss of either the Loss or PoteAtial Loss of BA)' two barriers.
Loss of aA)' two barriers and Loss or . Fuel Clad or RCS barrier. . . PoteAtial Loss of the third barrier. Fuel Clad Barrier RCS Barrier Containment Barder LOSS POTENTIAL LOSS POTENTIAL LOSS POTENTIA.L I:,QSS I:,QSS wss 1. RCS Aetivity 1. Primer,* Containment Pressure 1. Primar,* Containment Conditions A )1:. ESite s13eeifie Not A1313lieable A rc. PriHrnry l'fot Applieable A Yi. Yl'~Pbi\-l-R>rE;9
<l 2 r. Primary iAdieatioAs that eoRtaiAmeAt rapid drop iA eoAtaiAmeAt reaetor eoolaAt press1:1re greater primary press1:1re greater aeti,.,ity is greater tl,aA Esite speeifie eoAtaiAmeAt thaA Esite thaA 3QQ ):l:Ci,lgm val1:1e) d1:1e to RCS 13ressure followiAg s13eeifie value) dose equivaleAt I leakage. 13rimary OR rn+. eoAtaimneAt B. Esite pressure rise speeifie OR e~fplosi'te B. Primafj' mi~fture) e~tist s eoRtaiAmeRt iAside 13rimary pressure respoAse eoAtaiAmeAt Rot eoAsisteAt
'vvith OR LOCA eoAditioAs.
C. HC+L e*eeeded.
: 2. RP&#xa5; ~later Le*,el 2. RP&#xa5; Water Level 2. RP&#xa5; Water Le*,el <l 2L Primafj' A rr. :R:..r.H,l water A 1t. RPV water tfot A1313lieable Not A1313lieable A 11:. Primary eoAtaiAmeAt level eaAAot be level eaAAot be eoAtaiAment floodiRg required. restored aAd Festered and floodiAg maiAtaineEI abo','e maiAtaiRed above requires.
Esite speeifie R,.0 V Esite s13eeifie RPV water level water level 1 : __ --+ ..... --*-.J'.._ -_..._., L(_.._,....,.._,[lU l f,; \.V *-u .-....... ,., ........ 112 
[) .. . ] . i .
* 11-. -~* 
 
Basis Iefurmatien Fer BWR EAL Fissien Preduet Berrier Table 9 F 2 l>J E I 99 0 I (Re~*isieR 6) l>J eveFRSer 2012 BWR FUEL CLAD BARRIER THRESHOLDS:
The fuel Clad barrier coAsists of the zirealloy or stainless steel fuel bundle tubes that contain the fuel pellets. 1. 2. RCS Aetivity Loss 1.A This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;Ci/gm dose equivalent I 131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an apprmdmate range of2% to 5% fuel clad damage. Since this condition indicates that a sigAificant amount of fuel clad damage has occurred , it represents a l oss of the fuel Clad Barrier. There is AO Potential Loss threshold associated with RCS Acti:vity. Denleper Netes: Threshold values should be determined assuming RCS radioacfr,ity concentration equals 300 &#xb5;Ci/gm dose equivaleAt I 131. Other site specific units ma)' be used (e.g., &#xb5;Ci/cc). Depending upon site specific capabilities , thi s threshold may have a sample aAalysis component and/or a radiation monitor reading component.
Add this paragraph (or simi l ar wording) to the Basis if the threshold includes a sample analysis component , " It is recognized that sample collection and analysis of reactor coolant 1,yith highly elevated actiYity leYels could require several hours to complete. Nonetheles s , a sample related threshold is included as a backup to other indications." RPV *water Level Loss 2.A The Loss threshold represents the EOP requirement for primary containment flooding.
This is identified in the BWROG EPGs/SAGs when the phrase, " Primary Containment flooding ls Required ," appears. Since a site specific RPV water leve l is not specified here , the Lo s s threshold phrase, " Primary containment flooding required ," also accommodates the EOP need to flood the primary containment 1 ,1,'hen RPV water level cannot be determined and core damage due to inadequate core cooling is belie 1 ,red to be occurring. Potential Loss 2.A This water level corre s ponds to the top of the active fuel and is used in the E OPs to indicate a challenge to core cooling. 115 BWR FUEL CLAD BARRIER THRESHOLDS:
N E I 99 0 I (RevisieA
: 6) Jl,levember 2012 The R..0 V *water level threshold is the same as RCS barrier Loss threshold 2.A. Thus , this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergenC)' classification level to a Site Area Emergency. This threshold is considered to be exceeded 1.vhen , as specified in the site specific EOPs , RPV water cannot be restored and maintained above the specified level following depressurization of the R..0 V (either manually , automatically or by failure of the RCS barrier) or when procedural guidance or a lack of lov,r pressure R.0 V injection sources preclude Emergency R..0 V depre s surization.
EOPs allow the operator a wide choice of R..0 V injection sources to consider when restoring RPV water level to *within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low pressure injection sources. In some events , elevated R..0 V pressure may prevent restoration ofRPV 'Nater leYel until pressure drops helow the shutoff heads of available injection sources. Therefore , this Fuel Clad barrier Potential Loss is met only after either: 1) the R..0 V has been depre s surized , or required emergency R.&deg;'/ depressurization has been attempted , gh,ring the operator an opportunity to assess the capability of low pressure injection sources to restore R..0 V 1 Nater level or 2) no low pressure RPV injection systems are available , precluding RPV depressurization in an attempt to minin1ize loss of RPV inventory.
The term " cannot be restored and maintained above" means the value of RPV water level is flOt able to be brought above the specified limit (top of active fuel). The detenniflation requires an evaluation of system performance and availability in relation to the RPV 1.vater level 1 ,alue and trend. A threshold prescribing declaration when a threshold value etmnet he restored and maintained above a specified limit does not require immediate action s imply because the current value is belmv the top of active fuel , but does not permit extended operatiofl below the limit; the threshold must he considered reached as sooA as it is apparent that the top of active tl:lel cannot be attaiAed.
In high power i\TWS/failure to scram events , EOPs may direct the operator to deliherate l y lov,'er RPV water level to the top of active fuel iA order to reduce reactor power. RPV water level is then controlled betweeA the top of active fuel and the Minimum Steam Cooling R.0 V Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier , the immediate need to reduce reactor power is the higher priority.
For such eveAts , ICs SAS or SS5 will dictate the need for emergency classification. Since the loss of ability to determiAe if adequate core cooling is being provided presents a significant challeAge to the fuel clad barrier , a potential loss of the fuel clad barrier is specified.
116 
}I J;I 99 0 I (Revi s i e A 6) }l evember 2012 BWR FUEL CLAD BARRIER THRESHOLDS:
: 3. 4. DevelepeF Netes: Loss 2.A The phrase, " Primary containment flooding required ," should be modified to agree *.vith the site specific EOP phrase indicating eM,it from all EOPs and entry to the SAGs (e.g., drywell flooding required , etc.). Potent i al Loss 2.A The decision that "RPV \Yater level cannot be determined" is directed by guidance given in the RPV *water level control sections of the EOPs. Net Applieable (ineluded feF eumbeFieg eeesisteney between banieF tables) PFimary Ceetaiemeet Radiatiee Loss 1.A The radiation.
monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment , assuming that reactor coo l ant activity equals 300 &#xb5;Ci/gm dose equivalent I 131. Reactor coolant activity above this level is greater than. that e1tpected for iodine spikes and corresponds to an apprmdmate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred , it represents a l oss of the Fuel Clad Barrier. The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 1.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. 1-lote that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.
There is no Potential Loss threshold associated 1 Nith Primary Containment Radiation.
DevelepeF Netes: The reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory , *,*,ith RCS radioactivity concentration equal to 300 &#xb5;Ci/gm dose equivalent I 131 , into the primary containment atmosphere. 117 1'1e l 99 O I (Revi s ieR 6) 1'1e veffieer 2012 BWR FUEL CLAD BARRIER THRESHOLDS:
: 5. 6. Other Iedieatioes Loss and/or Potential Loss 5.A . . h Id th t , ee incl1:1ded to indicate This s1:1ecategory addresses other site specific thres o sa ma?-. loss or potential loss of the F1:1el Clad earr~er e~sed on plant specific design characteristics not considered in the genenc gmdance. De&#xa5;elof)er Notes: Loss and/or Potential Loss 5.A Developers sho1:1ld determine if other reliaele indicators e1d_st t; ~va~1:1at~
t~ st~t~:~;;his fission product earrier (e.g., review accident analyses descn~e mt st e t?; ~nal *sis Report as 1:1pdated). The goal is to identify any 1:!A1q1:1e or ~1te spec, c ~;dic~tio;;
that ~Yill promote timely and accl:!fate assessment of earner status. Any added thresholds sho1:1ld represent apprmdma~el?'
the san:ie r:lat~:e t~r!~:olds earrier as the other thresholds in this col1:1mn.
Basis mformat1on
<<->r e o may be used to gauge the relative barrier threat level. Emergeeey Direetor Judgment Loss 6.A b d b , the Emergency Director This threshold addresses any other factors ~ha~ are toe l:!Se) in determining whether the F1:1el Clad Barner is lost. Potential Loss 6.A b d b , the Emergency Director in ::~s~:==:=!:*::~
!~~.~~:~:~~e:~:fa:::i,~+.:i*:
~::::~ barrier status cannot ee monitored. Develof)er Notes: 118 
~1g 1 9 9 01 (Re~*i s i e A 6) ~IS V0ffiB0f 2 0 12 BWR RCS BARRIER THRESHOLDS:
The RCS Barrier is the reactor coolant sy s tem pressure boundary and includes the RPV and all reactor coolant system piping up to and including the isolation 1 ,*ai'les. 1. 2. Primary Centoinment Pressure Loss 1 .A I The (site specific value) primary containment pressure is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the EGGS or equiva l ent makeup system. There is no Potential Loss threshold associated with Primary Containment Pressure.
DeYeleper Netes: RPV Weter LeYel Loss 2.A This water level corresponds to the top of active fuel and is u s ed in the EOPs to indicate challenge to core cooling. The R.UV water JeyeJ threshold is the same a s Fuel Clad barrier Potential Loss threshold 2.A. Thus , this threshold indicate s a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is con s idered to be e>weeded 1 Nhen , as specified in the site specific EOPs , RPV \Yater cannot be restored and maintained above the specified level follov,*ing depressurization of the R.UV (either manually , automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depre ss urization EOPs allow the operator a wide choice of RPV injection sources to consid e r when restoring RPV wat e r l e 1 ,r e l to within prescribed limits. EOPs also specify depressurization of the R.0 V in order to facilitate R.&deg;'/ water level control with low pressure injection sources. In some eyents , elevated RPV pressure may prevent restoration of R.UV water level until pressure drop s belov,' the shutoff heads ofayailable injection sources. Therefore , this RCS barrier Lo ss i s met only after either: 1) the RPV ha s been depressurized , or required emergency RPV depressurization has been attempted , giving the operator an opportunity to assess the capability of low pressure injection sources to restore R.&deg;'/ 'Nater level or 2) no lov, pre s sure R.UV injection systems are available , precluding RPV depressurization in an attempt to minimize loss of R.UV inventory.
119 1-n;1 99 01 (Revi s ion 6) 1-Joyeffieer 2012 BWR RC8 BARRIER THRE8HOLD8:
: 3. The term, " cannot be restored and maintained above ," means the value ofR..0 V 1.vater level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of S)'Stem performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value ce,mot be restored and maintained above a specified limit does not require immediate action simply because the Ctlffent value is below the top of active fuel , but does not permit e1ttended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
In high power AT\VS/failure to scram events , EOPs may direct the operator to deliberately lov,rer RPV water level to the top of active fuel in order to reduce reactor power. R..0 V 'Nater level is then controlled between the top of active fuel and the Minimum Steam Cooling R..0 V Water Level (MSCRWL).
Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.
For such events , ICs 8A5 or 885 will dictate the need for emergency classification. There is no RCS Potential Loss threshold essoeiated with RPV Water Level. RCS Leek Rote Loss Threshola 3.A Large high energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure retaining capability of the RCS until they' ere isolated.
If it is determined that the ruptured line cannot be promptly isolatea from the Control Room , the RCS barrier Loss threshola is met. Loss Threshold 3.B Emergency RPV Depressurization in accordance
*with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed , the plant operators are directed to open safety relief 1 ,1elves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool , a Loss of the RCS barrier e1dsts due to the diminished effectiveness of the RCS to retain fission products *within its boundary.
Potential Loss Threshold 3.A Potential loss of RCS based on primary system leakage outsiae the primary' containment is determined from EOP temperature or radiation Mrur Normal Operating values in areas such as main steam line tunnel, RCIC , HPCI , etc., 1.vhich indicate a direct path from the RCS to areas outside primary containment.
A Max Normal Operating 1 ,1alue is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.
1 20 l>I E I 99 01 (Re~*ision
: 6) l>loveA'lber 2012 BWR RCS BARRIER THRESHOLDS:
: 4. The indicators reachiHg the threshold barriers and confirmed to be caused by RCS leakage from a primary system 1 Narrant an Alert classification.
A primary system is defined to be the pipes , valves , and other equipment which connect d i rectly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the S)'Stem. An UNISOLABLE leak which is indicated by MaJt ~formal Operating values escalates to a Site Area EmergenC)'
when combined 1.Yith Containment Barrier Loss threshold 3.A (after a containment isolation) and a General EmergenC)'
when the Fuel Clad Barrier criteria is also e,weeded.
DeYel0f)eF Netes: Loss Threshold 3.A The list of systems included in this threshold should be the high energy lines which, if ruptured and remain unisolated, can rapidly depressurize the RPV. These lines are typical I)' isolated b)' actuation of the Leak Detection system. Large high energy line breaks such as Main Steam Line (MSL), High Pressure Coolant Injection (HPCI), Feedwater, Reactor Water Cleanup (RWCU), Isolation Condenser (IC) or Reactor Core Isolation Cooling (RCIC) that are illHSOLABLE represent a significant loss of the RCS barrier. PFimaey Ceetaiemeet Radiatiee Loss 4.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment , assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 4 .A since it indicates a loss of the RCS Barrier only. There is no Potential Loss threshold associated with Primar)' Containment Radiation.
The reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory, 1 Nith RCS activity at Technical Specification allowable limits , into the primary containment atmosphere.
Using RCS activity at Technical Specification allowable limits aligns this threshold
'>Yith IC SU3. Also, RCS activity at this level will typically result in primary containment radiation levels that can be more readily detected by primary containment radiation monitors, and more readily differentiated from those caused by piping or component "shine" sources. If desired, a plant ma)' use a lesser 1 ,alue of RCS activity for determining this value. 121 l>JE I 99 0 I (R i: wi s i e A 6) l'l9 Y9fA09F 2012 BWR RCS BARRIER THRESHOLDS:
: s. 6. In some cases , the site specific physical location and sensitivity of the primary containment radiation monitor(s) ma)' be such that radiation from a cloud of released RCS gases cannot be distinguished from radiation emanating from piping and components containing elevated reactor coolant activity.
If so , refer to the Developer Guidance for Lo ss/Potential Loss 5.A and determine if an alternate indication i s available. Other Iedieati0es Loss and/or Potential Loss 5.A This subcategory addresses other site specific thresholds that may be included to indicate loss or potential lo s s of the RCS barrier based on plant specific design characteristics not considered in the generic guidance.
Denl0per N0tes: Loss and/or Potential Loss 5.A Developers should determine if other reliable indicators e,dst to evaluate the status of this fission product barrier (e.g., review accident analyses described in the site Final Safety .Analysis Report , a s updated).
The goal is to ideRtify any unique or site specific indicatioRs that 1 Nill promote timely aRd accurate assessmeRt of barrier starus. Any added thre s holds should represent apprmdmately the same relative threat to the barrier as the other thresholds in this column. Basis information for the other thresholds ma)' be used to g auge the relative barrier threat level. Emergeeey Direet0r Judgment Loss 6.A This threshold addresses any other factors that are to be used by the E mergency Director in determining whether the RCS barrier is lost. Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider v,'hether or not to declare the barrier poteRtially lost in the event that barrier status cannot be moRitored.
Devel0per N0tes: 122 BWR CONTAINMENT BARRJFU >lSl 99 QI E~ev;,;,**l r:a:n: THRESHOLDS
~le~*e mlm 2 0 1 2 T=h~e:P:r:i1:=H:aP=*~
'C ;;o;~;* ;;::;::::::!:~;;;:=
00-"'."'twoll,,theif,,.,""'""''
"*
* J n amment Ba . . ~ntereonneeting athrner meludes the dr a" ~eloti~R valve/ce:;.~::::
eeRReetieR s up 1~ *~=~*;!~~ :etwell , their respeetive om nlert to a Site Area B amer thresholds are used " iRg the outellHest eeRtai 1. n . mergeRey or a GeReral B merge~e~nto,1a fer esealotieR ef 1:'~~L r FIBl&l'y C I . r t ee tuemeet Ceeditiees boss 1.A and 1.B IS m,pe-t heea*se it i;:~:*i~~~.".!!s~g,,ed.
The ** .:i,"::.".!:dc:':'.'!:
fer the eoHditioR PoteRlial Less 1 , r -er a eoR1a1R1HOFII hypass eeRdi;1.'
Jl,JED) response ., , on. :::: :::::::R:=s~:~:t i~Rr:.!~:::7R!::1:::::~==
;R;::::s .: a n mtegnty er t ese eonditio . d. n. us ,
* ns m 1eates a I These thresholds I oss of d re yon ope t ~n: therefore a speeifie.. r~ or reeognition of an un los f . ensat1on effi t ) pressure (1 e s e pnmary eeRtaiRffle I. ---ee+ fellowiRg aR iRiti 1* ., Rel attrihutahle to result efmass and e R iRlegnty.
Prima,' ee . a pressure iaerease iad' Rapid U1'~PL'\N1'ffiD l . drywell spray or eond oss ?f pnmary eontainment The threshold pressure is . aeeeptanee testing demon!::a pnmary eontainment internal . If hydrogen eoneentratio plant EOPs , in an OK , n ~eaehes or en.seeds the 1011 The Heat Capaeity Te temperature from whi=p~rature Limit EHCTL) is the h' h mergeney RP" De . ,g est suppression p I
* g . ' pressHmaf , ..
* ee uppresS1eR ehamh !OR will *et raise* erte!'Hpe B
* suppression eha bra dre above the I * . m er and eq
* 41 ruu!'Hum te req*1red te eperote whea t!,e";!';~RI withiR the SHl'flFOS;;.:p:atu:
e"l'*_hility eftlie OR IS press*ri,ed , ""' er wh1eh may he 1 23 1'lel 99 01 (RevisieA 6) 1'JeveFA06F 2012 RWR CONT,A .. INMENT BARRIER THRESHOLDS:
: 2.
* Suppression chamber pressure above Primary Containment Pressure Limit l\, \Yhile the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent. The HCTL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore , the inability to maintain plant parameters belov,r the limit constitutes a potential loss of containment.
De&#xa5;el0per Netes: Potential Loss l .B BWR EPGs/8AGs specifically define the limits associated with explosive mi><.tures in terms of deflagration coHceHtratioHs of hydrogen and m(ygen. For Mk I/II coHtainmeHts the deflagratioH limits are " 6% hydrogen aHd 5% m(ygeH iH the dryv,rell or suppressioH chamber". For Mk III containments , the limit is the " Hydrogen Deflagration Qyerpressure Limit". The threshold term " e}(plosive mixture" is synonymous with the EPG/8AG " deflagration limits". Potential Loss l .C Since the HCTL i s defined assuming a range of suppression pool *Nater levels as low as the elevation of the downcomer openings iH M]c I/II containments , or 2 feet above the elevation of the horizontal vents in a Mk III coHtainment , it is um1ecessary to consider separate Containment barrier Loss or Potential Loss thresholds for abHormal suppressioH pool water level conditions.
If desired , deYelopers may include a separate Containment Potential Loss thfeshold based on the inability to maintain suppressioH pool water level abo,*e the dovmcomer openiHgs in Mk I/II containments , or 2 feet above the elevation of the horizontal vents in a Mk Ill containment 1 Nith RPV pressure above the minimum decay heat removal pressure, if it will simplify the assessment of the suppression pool level con~ponent of the HCTL. RPV Water LeYel There is no Loss threshold associated with RPV Water Level. Potential Loss 2.A The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold 2.A. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be restored and maintained and that core damage is possible.
BWR EPGs/SAGs specify the conditions that require primary containrneHt flooding. '.VheH primary coHtainmeHt floodiHg is required , the EPGs are e}dted and SAGs are entered. Entry iflto SAGs is a logical escalation in response to the inability to restore and maiHtain adequate core cooliHg. 124 
~!B l 99 01 (ReYi s i e A 6) ~le vemser 2012 BWR CONTi. .. INMENT BARRIER THRESHOLDS:
: 3. PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which , if not corrected , could lead to RPV failure and increased poteHtial for primary containment failure. lA conjunction with the RPV *water level Loss threshold s in the Fuel Clad and RCS barrier columns , this threshold results in the declaration of a General Emergency.
De&#xa5;el01Jer Netes: The phrase, " Primary containment flooding required ," should be modified to agree 1 Nith the site specific EOP phrase indicating e x it from all EOPs and entry to the SAGs (e.g., dr)0 , 1 ,'ell flooding required , etc.). Primary Centainmeet Iselatiee Failure These thresholds address incomplete containment i s olation that allov,s an UNISOLABLE direct relea s e to the environment.
Loss 3./\ The use of the modifier " direct" in definin g the release path discriminates against release path s through interfa c ing liquid sy s tems or minor release pathways , such a s in s trument lines , not protected b y the Primary Containment I s olation S)'Stem (PCIS). The e 1 dstence of a filter is not considered in the threshold assessment.
Filters do not remove fission produ c t noble ga s es. In addition , a filter could become ineffective due to iodine and/or particulate loading beyond de s ign limits (i.e., retention ability has seen exceeded) or water s aturation from steam/hi g h humidity in the release stream. follO'tYing the leaka g e of RCS ma s s into primary containment and a rise in primary containment pressure , there may be minor radiological releases associated with allov , aele primary containment leakage through variou s penetrations or system components.
Minor releases may also o c cur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category /\ !Cs. Loss 3.B EOP s may direct primary containment isolation valve logic(s) to be intentionally bypassed , even if offsite radioactivity release rate limits will be e1rneeded.
Under these conditions with a valid primary containment isolation signal , the coHtainment should also be considered lost if primary containment venting is actually performed.
Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment.
Venting for primary containment pressure control 1when not in an accident situation (e.g., to control pressure belO'tv the drywell high pressure scram setpoint) does not meet the threshold condition.
125 Loss 3.C N E! 99 O I (Re,*isieH 6) *Novembe r 2012 The Mme Safe Operating Temperature and the Ma1, Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant ,viii fail , nor (2) personnel access necessar)'
for the safe shutdown of the plant will be precluded.
EOPs utilize these temperatures and radiation levels to establish conditions under which R.DV depressurization is required.
BWR CONTAINMENT BARRIER THRESHOLDS:
: 4. The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes , valves , and other equipment 1 Nhich con:nect directly to the RPV such that a reduction in RPV pressure ,vill effect a decrease in the steam or water bei:ng discharged through an unisolated break in the system. In combination 1.vith RCS potential loss 3.A this threshold would result in a Site Area Emergency.
There is no Potential Loss threshold associated
*with Primary Containment Isolation Failure. Develaper Nates: Loss 3.B Consideration may be given to specifying the specific procedural step within the Primary Containment Control EOP that defines intentional venting of the Primary Containment regardless of of:fsite radioactivity release rate. Primary Cantainment Radiatian There is no Loss threshold associated with Primary Containment Radiation.
Potential Loss 4 .A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.
NUREG 1228 , Source Estimtltiens During Incident Rcspensc te Sc.rcrc 1 Vuclctlr Pewcr Plant Accidents , indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to e1dst , there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency. 126 DeYeleper Netes: N E I 99 0 I (Re v isieA 6) J>l 0Y0ffi00F 20 J 2 " )IT, 1
* v,,,fer . r *e1, f Re-9pense fe oe>,rere , r'l;tetem , v' NUREG 1228 , S ource EshHu::1he~s
:urm~ i1~:t fu~l classing failure value. Unless . s th basis er usmg ~-e -Ji) Pl:tmfAccidc19fs , previ es~: .: s*fferent "alue the reasing sheuls be there is a site specific ana!ys1s Justifymg I s s'ispe;sal ef the reacter ceelant neble ,..I
* s ssuming the mstantaneeus re ease an . tfetermme a : . s ... *th 200~ fuel clas failure inte the pnmary gas ans iesine mventery assec1ate vvl ' centainment atrnesphere.
llWR CONTAINMENT BARRIER THRESHOLDS:
: 5. 6. Other Indieetiens DeYeleper Netes: Less ans/er Potential Less 5.A . . r ble insicaters e1dst te evaluate the status ef this Develepers sheuls setermme if _ether r~~a t lyses sescribes in the site Final Safety b
* E re"Ie"' acc1-en* ana fissien presuctarnere.g., ~h" 1
* te isentify any unique er site specific A I rsis R::epert as upsates).
e gea 15 f b
* t t nna ) ' . . I , s accurate assessment earner s a us. insicatiens that will promete time Yan imatel , the same relative threat te t he Any asses threshelss sheuls r~pre~ent apprm. B . ? t; matien for the ether thresholss barrier as the ether threshelss m this celumn. as1s m~r may be uses te gauge the relative barrier threat le,vel. Emergeney Direeter Judgment Less 6.A b s b r the emergency Directer This thi=eshels assresses any ether facters th~t a~e tee use) in setermining whether the Centainment barner is lost. Potential Less 6.A , b uses by the Emergency Directer in This threshels assresses any et_her facters t_hat_ma~te:tially le~t. The emergency setermining whether the ?entm:m:nt Bam::;scieclare the barrier petentiall)
' lest in the Directer sheuls alse cens1ser w et er ?r ne event that barrier status cannet be men1teres.
DeYeleper Netes:
I 1. A. }/El 99 01 (Re,*isioA
: 6) }lovember 2012 Table 9-F 4 B-: :PWR-EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FAlALERT Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier. OQerating Mode AQQlicabili!Y:
1: 2: 3: 4 Fuel Clad Barrier LOSS POTENTIAL LOSS Critical Safetv Function Status 1. Conditions A. Conditions reguiring entry reguiring entry into into Core Cooling Core Cooling RED Path (CSP ORANGE Path C. l) are met. (CSP C.2) are met. OR B. Conditions reguiring entr)'.'.
into Heat Sink RED Path (CSP H. l) are met. FSl SITE AREA EMERGENCY FGl GENERAL EMERGENCY Loss or Potential Loss of any two barriers.
Loss of any two barriers and Loss or Potential Loss of the third barrier. OQerating Mode AQQlicabili!Y:
l = 2: 3: 4 OQerating Mode AQQlicabilitv:
l: 2: 3. 4 RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS Critical Safe!l'. Function Status 1. Critical Safetv Function Status Not A1;mlicable A. Conditions Not Agglicable A. I. Conditions reguiring entry into reguiring Heat Sink RED ent!}'.Entrv into Path (CSP H. l) are Core Cooling met. RED Path OR (CSP C. l) are B. Conditions met. reguiring entry into AND RCS lntegri!)'.'.
: 2. CSP C.1 not RED Path (CSP effective within P. l) are met. 15 minutes. 12. RCS or SG Tube Leakage -12. RCS or SG Tube Leakage -12. RCS or SG Tube Leakage Not Applicable Not AQQlicable A. An automatic or A. Operation of a A. A leaking or Not Applicable 6 ] t. RGS#eaeteF manual ECCS (SI) standby charging RUPTURED SG is ,,*essel te,,*el less actuation i s required (makeup) pump is FAULTE D outside of H1an Esite s13eeifie by EITHER of the required by containment.
le&#xa5;el~.Genaitiens following:
EITHER of the Fe&sect;HiFing eni!:)* inte 1. UNISOLABLE following:
GeFe Geeling RCS leakage 1. UNISOLABLE GRA}mE Path OR RCS leak age (GSP G.2) Elfe met +/-:-SG tube OR RUPTURE. 2--:--SG tube 3-s-leakage. 128 Fuel Clad Barrier LOSS POTENTIAL LOSS LOSS +.2-,_ RCS Barrier POTENTIAL LOSS L OR B. R:GS eeelde't 1 rA Fate g&#xa5;eateF thaA Esite s13eeifie f3feSSt1Fi~ed theffflal sheek eFiteFiall imits defiAed by site s13eeifie iHdieatieAs).Ge AditieHs f0&sect;ttiFiAg 0At!:Y iAte R:GS IAtegFifr REE} Patl~ EGSP P. l 1 aFe met 129 Ne:! 99 01 (Re&#xa5;isieR
: 6) }levemeer 2012 Containment Barrier LOSS POTENTIAL LOSS Fuel C l a d Barr i er LOSS POTENTIAL LOSS I ,, T _ _. -yy __ .._ n 1 -* ;--*-.. -* A r r. GeFe eM::it 6 l r. GeFe e~Ht tAeFffieee1:1ple tAeFffieee1:1ple Feaaings gFeater reaaings gFeateF tAal'I (site speeifie tAan (site speeifie teffiperaturn ten:iperature
., , a11:1e).Genaitie1:is val1:1e).Genaitien s req1:1iFing entFy req1:1iFing ent5 1 inte inte GeFe Geeling GeFe Geeling REE> PatA (GSP ORA1'JGE Path G.1) are ffiet. (GSP G.2) are ffiet. OR B. lnaaeEj1:1ate RGS Aeat Feffie*,*al eapabili~* 't 1 ia s teaffi generateF s a s i nai e atea by (s it e s peeifie inaieatien s).Genait ien s Feq1:1iring entF)' i1:ite Heat Sinl E R:&#xa3;9 PatA (GSP H. l) aFe met. --RCS Barrier LOSS POTENTIAL LOSS .., y __ _. ,.. -a-n .1 -* -.. . -* 1'let Appli e able A rr. h=iaaeEjl:late
-RGS Aeat Feffie1t*al eapabilt~* *,*ia steaffi geRerateFs as iRaieatea O)' (s ite speeifie inaieatien s).Genaiti ens req1:1iFing entry inte Heat Sink RB9 PatA (GSP f J.1) aFe ffiet. -130 1-JE I 99 01 (Revisi o A 6) No*,*e1'Aser 20 12 Containment Barrier LOSS POTENTIAL LOSS .., y __ ... ..__ yy __ .._ n I -* *-. -* 1'Jet Applieable A t*"t. I. (Site -s peeifie eFiteFia feF -entry inte ee Fe eeelin g FesteFatien
-preeea1:1re)EntFy inte Gere Geeling RED PatA (GSP G.1) aFe ffiet. ,\:.""TD ,.., -* Resteratien flFeeea1:1Fe net e#e eti&#xa5;e witAin l 5 ffi i n1:1 t e s.G S P G.I net e f feefr t'e witi'lin 15 rn in1:1te s.
JJ.. RCS Activity/
Containment Radiation I L Containment Not Applicable radiation monitor reading greater than 577 R l hr indicated on ANY of the following.
* 1 (2)-RE-126
* 1 (2)-RE-127
* 1 (2)-RE-128 OR B. 1(2) (Site s13eeiJie iAeieatieAs that reaeteF eeelaAt aeti*,*it)'
is gi=eater thaA 388 f::!:Gilgffl eese eE!tliYaleflt l ~RE-109 greater than 4,500 mR/hr -JJ.. RCS Activity/
Containment Radiation JJ.. A. Containment Not Applicable radiation monitor reading greater than 11 R/hr indicated on ANY of the following~~
* l (2:))-RE-126
* 1 (2)-RE-127
* 1(2)-RE-128 131 }lei 99 01 (RevisioR
: 6) }/0 1 ,'eFRSeF 2012 RCS Activity / Containment Radiation Not Applicable A. Containment radiation monitor reading greater than 18,500 R/hr indicated on ANY of the following.
* 1 (2)-RE-126
* 1 (2)-RE-127
* 1 (2)-RE-128 Esite s13eeifie
&#xa5;&#xa3;Htler, I _14. Containment Integrity or Bypass 44. Containment Integrity or Bypass Not A pplicable Not A pplicable Not Applicable Not Applicable I I I s. 9theF lndieatiaes
: s. 9theF lndieatiaes l>fot ApplieableA.
l>fot ApplieableA.
l>fot ,<\pplieableA.
l>let ApplieableA. (site Sf:leeifie (site Sf:leeifie as (site Sf:leeifie as (site Sf:leeifie as as af:lf:llieaele) af:lf:llieaele) af:lf:llieeele) af:)f:)lieaele) 132 }IE! 99 01 (ReYisieA
: 6) }le\1 emeer 20 12 ,14. Containment Integrity or Bypass A. Co ntainment i so l atio n A. Containment is requ ired pressure g reater -AND than (site Sf:leeifie EITHER of the ~60psig. following:
OR 1. Co ntainm e n t B. E11:f:llesi&#xa5;e mi1ttl::1Fe i nte grity has been e11:ists iAsiEle l ost based on eentainmeAt6%
H 2 Emergency inside containment.
Director OR judgment.
C. 1. Contai nmen t OR pressure greater 2. UNISOLA B LE th an B,-t-te-pathway from the Sf:leeifie containment to f:lF6SSt!Fe the e n viro nm ent setf:)eiAt)25 exists. Qfil.& OR AND B. Indications of RCS 2. Less than one leakage outside of fu ll train of containment. (site Sf:leeifie s:;*stetfl eF eq u i f:)ment)depr essurization eguipment is operating per design for 15 minutes or -l o n ger. s. 9theF ledieatians l>let ,<\pp! ieable,6 .. l>let ApplieaeleA. (site Sf:leeifie as (site Sf:leeifie as af:)f:)lieaele)
Bf:lf:llieeele)
I 6~. Emergency Director Judgment A. ANY condition in A. ANY condition in the opinion of the the opinion of the Emergency Emergency Director that Director that indicates Loss of indicates Potential the Fuel Clad Loss of the Fuel Barrier. Clad Barrier. 56. Emergency Director Judgment A. ANY condition in the A. ANY condition in opinion of the the opinion of the Emergency Director Emergency that indicates Loss of Director that the RCS Barrier. indicates Potential Loss of the RCS Barrier. 133 ~-A. l>Je.1 99 01 (R e;*isioA 6) J>I0&#xa5;6FA06F 20 J 2 Emergency Director Judgment ANY condition in the A. ANY condition in opinion of the the opinion of the Emergency Director Emergency Director that indicates Loss of that indicates the Containment Potential Loss of Barrier. the Containment Barrier.
Basis Information For PWR EAL Fission Product Barrier Table 9-F-! Dw1el0per N0tes: Thresh0ld Parameters and Values NEI 99 0 I (R ev i s ioA 6) J>Jov ember 2012 Each PWR owner's group ha s de 1 ,*eloped a methodology for guiding the development and implementation ofEOPs (i.e., assessing plant parameters, and determining and prioritizing operator actions). Many of the thresholds contained in the P'.llR EAL Fission Product Barrier Table reflect conditions that are specifically addressed in EOPs (e.g., a loss of heat removal capability by the steam generators).
'.Vhen developing a site specific threshold , developers should use the parameters and values specified within their EOPs that align *.vith the condition described by the generic threshold and basis , and related developer notes. This approach 'Nill ensure consistency betv,*een the site specific EOPs and emergency classification scheme , and thus facilitate more timely and accurate classification assessments. In support ofEOP development and implementation , the Westinghouse Ovmers Group (WOG) developed a defined set of Critical Safety Functions as part of their Emergency Response Guidelines.
The WOG approach structures EOPs to maintain and/or restore these Critical Safety Functions , and to do so in a prioritized and systematic manner. The WOG Critical Safety Functions are presented below.
* Subcriticality
* Core Cooling
* Heat Sink
* RCS Integrity
* Containment
* RCS Inventory The WOG ERGs pro 1 ,ride a methodology for moAitoring the status of the Critical Safety Functions and classif)*ing the significaAce of a challenge to a function; this methodology is referred to as the Critical Safety FunctioA Status Trees (CSFSTs). For plants that have implemeAted the WOG ERGs , the guidance in }~El 99 01 allov,rs for use of certaiA CSFST assessment results as EALs and fission product barrier loss/potential loss thresholds.
lH this manner , an emergeAcy classification as s essment may flov,r directly from a CSFST assessment. It is importaAt to uAderstand that the CSF8Ts are evaluated usiAg plant parameters , aAd that they are simply a vendor specific method for collectively evaluatiAg a set of parameters for purposes of driviAg emergenc:,* operating procedure usage. For the emergeAcy coAditions of iAterest , the geAeric thresholds withiA the P'NR EAL FissioA Product Barrier Table specif:,* the plant parameters that defiAe a potential loss or loss of a fission product barrier; hov,*ever , as described in the associated Developer Notes , a C8F8T terminus may be used as 'Nell. For this reason , iAclusion of the C8F8T related tlH=esholds would be redundant to the parameter based thresholds for plaAts that employ the WOG ER.Gs. Sites that employ the WOG ERGs may , at their discretioA , iAclude the C8F8T based loss and poteAtial loss thresholds as described in the Developer Notes. Developers at these sites should coAsult with their classification decision makers to determine if iAclusion would assist with timely and accurate emergeAcy classification.
This decisioA should consider the effects of aAy site specific changes to the generic WOG C8F8T evaluation logic and setpoints , as *,-.,,ell as those 134 NEI 99 O I (Rev isioA 62 ~IO\'e1H0er 20L . *ng from user rules ~Pp I d to specific accident con E e~weptions to , erating procedurese.g., ort system). !"cable to emergenc)
_op ditions or loss ofa supp ans1 , or trans1t1on He proee6ure eRtr; . f3 ways: ' Be a66,esse6 tR ORO o . Re De*,elope, Th CSFS+ thresholds ma) d "alues as discussed m t -r-Ne ameters an d* thresho l ds vrill use par 1) Not iROOfjlOFale-, ,.. 16 fiOYO 2 ' 8 fuel ela6 lo,s ""u,, -. , OH6 value tkreskol6s (e.g., Rt, . eoR6itioRs met . I orporated along with paral~~~~F" and " Core Cooling Red e ) 2) nc h " CE+s > :z::1:1 t L.:resholds sue as Id H for all thresho s. . amete,s aR6 values I, 16s tkeR all s u eA 3) Used in heu of par . I de the CSFS+ based thres o 1:, the C Orange
* ade to me u * " ble to use on J . *, eeeisioR ""' *t i s Rot P""""" 1 Re, CSFST ''litk oRe ei<eOftltoR, , Be u,ee ifl tke taBle (e.g., 'J, Role OH6 6isrega,e al ot Hke ' a tk eskol6s lflUS fi I la6 l,ar,1e, t ,.., SF8T Beeause o allo":'e-__ , a poteRtial loss oHhe :'" e the RCS !Rteg,ity (P) C ! ~,essure telflpefftlttre te,mm~s askol6s). Tke ORO "*septto~
:stkat relies""'"'
assess"':" :016 without the Beed te Basee t Fes f h p Ree 6eeist0fl polfl RCS pot<>Rtial loss t,res complemty o t e d"f on may be used as an curve, a p Red con I ~8F8+ based thresholds.
incorporate the other 135 NEI 99 0 I (Re~*ision 6) November 2012 PW&-FUEL CLAD BARRIER THRESHOLDS:
The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. 1. 1. Critical Safety Function Status Loss I .A This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant. Core Coo l ing -RED indicates significant superheating and core uncovery and is considered to indicate loss of the Fuel Clad Barrier. CSP-C. I is the Critical Safety Procedure that provides directions to restore core cooling. Potential Loss I .A This reading indicates temperatures within the core are sufficient to allow the onset of heat-induced cladding damage. Core Cooling -ORANGE indicates subcooling has been lost and that some clad damage may occur. CSP-C.2 is the Critical Safety Procedure that provides directions to restore adequate core cooling. Potential Loss l .B This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e .. loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. Verification of the EALs for CSP-H.l should include an assessment of if feed flow was reduced based on the actions performed for an uncontrolled depressurization of both steam generators.
If this is the case, then declaration requirements should not be considered to be met. This does not affect the time that the CSFST initially changed color and met the CSP entry conditions for potential EAL event classification.
Alternatively, if CSP-H.l was entered during a loss of coolant accident, it may be exited if secondary heat sink is not required based on RCS pressure less than non-faulted SIG pressure.
If this is the case, then declaration requirements should not be considered to be met. RCS aF SC Tuhe Lealrnge There is no Loss threshold associated with RCS or SG Tuee Leakage. Potential Loss I .A This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat induced cladding damage. Core Cooling ORt\1'tGE indicates subcooling has seen lost and that some clad damage may occur. CSP C.2 is the Critical Safety Procedure that provides directions to restore adequate core cooling. 136 DeYeleper Netes: Potential Loss I .A }/el 99 01 (ReYisioA
: 6) }fovember 2012 Enter the site spec i fic reactor vessel water level value(s) used by EOPs to identify a degraded core cooling condition (e.g., requires prompt restoration action). The reactor vessel level that corresponds to apprmrimately the top of active fuel may also be used. For plants that have implemented
'.Vestinghouse Owners Group Emergenc:Y Response Guidelines , enter the reactor vessel level(s) used for the Core Cooling Orange Pa t h (including dependencies upon the status ofRCPs, if applicable).
Westinghouse ERG Plants Developers should consider including a threshold the same as, or similar to, "Core Cooling Orange entry conditions met" in accordance with the guidance at the front of this section. In accordance with EOPs , there may be unusual accident conditions dur i ng *,vhich operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted. PWR FUEL CLAD Bl .. RRIER THRE8HOLD8:
Meeting this threshold results in a Site Area E mergency because this threshold is identical to RCS Barrier Potential Loss threshold 2.A; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system. In accordance with EOPs. there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted.
137 N E l 99 0 I (Revi s ioR 6) }foverHeer 2012 FUEL CLAD BARRIER THRESHOLDS:
: 2. RCS or SG Tube Leakage There are no L oss or Potential Loss thresholds associated with RCS or SG Tube Leakag e. Potential Loss I .A This reading indicates a reduction in reactor vessel water level sufficient to allov, the onset of heat induced cladding damage. Develef)eF Netes: Potential Loss I .A Enter the site specific reactor vessel water level value(s) used by EOPs to identify a degraded core cooling condition (e.g., requires prompt restoration action). The reactor vessel level that corresponds to apprm,imately the top of active fuel may also be used. For plants that have implemented Westinghouse Owners Group Emergency Response Guidelines , enter the reactor vessel level(s) used for the Core Cooling Orange Path (including dependencies upon the status ofRCPs, if applicable).
Westinghouse ERG Plants Developers should consider including a threshold the same as , or similar to, " Core Cooling Orange entr y conditions met" in accordance 1 Nith the guidance at the front of this section. In accordance 1.Nith EOPs , there may be unusual accident conditions during 1.vhich operators intentionally reduce the heat removal capability of the steam generators
; during these condition s, clas s ification using threshold is not warranted.
Develef)eF Netes: Some site specific E OPs and ,l or EOP user guidelines may establish decision making criteria concerning the number or other attributes of thermocouple readings necessary to drive actions (e.g., 5 CETs reading greater than 1 , 200 6 F is required before transitioning to an inadequate core cooling procedure). To maintain consistency with EOPs , these decision making criteria may be used in the core exit thermocouple reading thresholds. Loss 2.A Enter a site specific temperature value that corresponds to significant in core superheating of reactor coolant. 1 , 200 6 F may also be used. For plants that have implemented
\l/estinghouse Owners Group EmergencJ' Response Guidelines , enter the parameters and values used in the Core Cooling Red Path. Potential Loss 2.A Enter a site specific temperature value that corresponds to core conditions at the onset of heat induced cladding damage (e.g., the temperature allov,ing for the formation of superheated steam assuming that the RCS is intact). 7006&#xa5; may also be used. 138 J>IE I 99 0 1 (Re v i s i e A 6) J>le veH10er 2012 For plants that have implemented Westinghouse Owners Group Emergency Response Guidelines , enter the parameters and value s used in the Core Cooling Orange Path. Potential Loss 2.B Enter the site specific parameters and values that define an e>ctren~e challenge to t he ability to remove heat from the RCS via the steam generators.
These will t)*pically be parameter s and values that would require o perators to take prompt action to address thi s condition.
For plants that have implemented We s tinghouse Ovmers Group Emergency Re s ponse Guidelines , enter the parameters and value s used in the Heat Sink Red Path. Westinghouse E RG Plants As a loss indication , developers s hould c o n s ider including a threshold the same a s, or similar to, " Core Cooling Red entry conditions met" in accordance 1 Nith the guidance at the front of this section. PWR FUEL CLAD BARRIER THRESHOLDS:
As a potential loss indication , developer s should consider inclt1din g a threshold the same as , or similar to, " Core Cooling Oran g e entry conditions met" in accordance with the guidance at the front of this section. As a potential loss indication , developers s hould consider including a threshold the same as , or similar to, " Heat Sink Red entr y c o ndition s met" in accordance with the guidance at the front of this section.
FUEL CLAD BA:RRIER THRESHOLDS:
i.L_RCS Activity/
Containment Radiation Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activ i ty equals 300&#xb5;Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred , it represents a loss of the Fuel Clad Barrier. The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 3.A since it indicates a lo ss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.
Loss 3.B 139 NEI 99 0 I (Re~*isioll 6) }IoYeffleer 2012 This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred , it represents a loss of the Fuel Clad Barrier. There is no Potential Loss threshold associated with RCS Activity/ Containment Radiation.
DeYelef)er Netes: Loss 3.A The reading should be detennined assuming the instantaneous release and dispersal of the reactor coolant Hobie gas aHd iodiHe iHveHtory , v,ith RCS radioactivity conceHtratioH equal to 300 &#xb5;Ci/gm dose equivaleHt I 131, iHto the coHtainment atmosphere.
P~'R FUEL CLAD BARRIER THRESHOLDS:
Loss 3.B Threshold 1 1alues should be determiHed assumiHg RCS radioactivity coHceHtratioH equa l s 300 &#xb5;Ci/gm dose equivaleHt I 131. Other site specific uHits may be used (e.g., &#xb5;Ci/cc). DepeHdiHg upoH site specific capabilities, this threshold may have a sample aHalysis compoHent aHd/or a radiatioH moHitor readiHg compoHent.
Add this paragraph (or similar wordiHg) to the Basis if the threshold iHcludes a sample analysis compoHeHt , "It is recogHized that sample collectioH aHd aHalysis of reactor coolaHt with highly elevated activity levels could require several hours to complete.
l>foHetheless, a sample related threshold is iHcluded as a backup to other indicatioHs." M~Containment Integrity or Bypass 4. Not Applicable (included for numbering consistency)
Other lndieatiens Loss aHd/or Potential Loss 5.A This subcategory addresses other site specific thresholds that may be included to iHdicate loss or potential loss of the Fuel Clad barrier based OH plant specific desigH characteristics Hot coHsidered iH the geHeric guidaHce.
Loss aHd/or Potential Loss 5 .A 140 
}lE I 99 0 1 (Rev i sieA 6) }Je v erf!ae r 2012 Developers s h o uld determine if other reliable indicator s e>dst to evaluate the statu s of this fission product barrier (e.g., revie\v accident analyses described in the site Final S afety /\nalysis Report , as updated).
The goal i s to identify an y unique or s ite specifi c indications that will promote timely and a c curate a s sessment of barrier status. Any added thresholds should represent apprmdmately the same relative threat to the barrier as the other thre s hold s in this column. Basis information for the other thresholds may be used to g au g e the relative barrier threat l e vel. 5. Emergency Director Judgment Loss 56.A This threshold addresses any other factors that may be used by the E mergency Director in determining whether the Fuel Clad Barrier is lost. P\&#xa5;R FUEL CLAD BARRIER THRESHOLDS:
Potential Loss 56.A This threshold addresses any other factors that may be used by the E mergency Director in determining whether the Fuel Clad Barrier is potentiall y lost. The E mergenc y Director should also consider whether or not to declare the barrier potentiall y lo st in the event that barrier status cannot be monitored.
DeYeleper Netes: 141 
}[ET 99 0 I (Re&#xa5;isioA
: 6) }foYet'Aser 2012 PWR-RCS BARRIER THRESHOLDS:
The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves , and other connections up to and including the primary isolation valves. 1. Critical Safety Function Status There is no Loss threshold associated with Critical Safety Function Status. Potential Loss I .A This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e .. loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. If CSP-H. l was entered during a loss of coolant accident, it may be exited if secondary heat sink is not required based on RCS pressure less than non-faulted SIG pressure. If this is the case, then declaration requirements should not be considered to be met. Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold l .B; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system. CSP-H.l is the Critical Safety Procedure that provides directions if the ultimate heat sink function is under extreme challenge.
This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature and thus a Potential Loss of the RCS barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted.
Potential Loss l .B This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock -a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e .. hot and pressurized).
CSP-P.l is the Critical Safety Procedure that provides directions to avoid. or limit thermal shock or pressurized thermal shock to the reactor pressure vessel or overpressurization conditions at low temperatures.
IF CSP-P. l is entered during a large break loss of coolant accident, it may be exited if RCS pressure is low enough to allow for RHR forward flow. If CSP-P.l is exited for this reason, then the Potential Loss criteria should not be considered met. 142 RCS BARRIER THRESHOLDS:
l.L_RCS or SG Tube Leakage Loss+2.A NE l 99 0 I (Rev i s i e A 6) J>levember 2 0 1 2 This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier. This threshold is app li cable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location -inside containment , to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to r equire a safety injection is considered to be RUPTURED.
If a RUPTURED steam generator is also FAUL TED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold l .A will also be met. Potential Loss -l-2.A This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI) actuation has not occurred. The threshold is met when an operating procedure , or operating crew supervision , directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer level. This threshold is applicable to unidentified and pressure boundary leakage , as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location -inside containment , to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
If a leaking steam generator is also FAULTED outside of containment , the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold l .A will also be met. Potential Loss 1.B This condition indicates an e J (treme challenge to the integrit)* of the RC8 pressure boundary due to pressurized thermal shook a transient that oauses rapid RC8 oooldown v,rhile the RC8 is in Mode 3 or higher (i.e., hot and pressurized).
C8P P. l is the Critical Safety Procedure that provides directions to avoid, or limit. thermal shook or pressurized therma l shook to the reactor pressure vessel or overpressurization conditions at lov,* temperatures. 143 
}JEI 99 0 I (RevisioA
: 6) }loverflser 2012 PWR RCS BARRIER THRESHOLDS:
: 2. DeYele~er Netes: Loss 1.A Actuation of the EGGS may also be referred to as Safety Injection (SI) actuation or other appropriate site specific term. Potential Loss 1.A Depending upon charging pump flov,r capacities and RCS volume control parameters, developers may use an RCS leak rate value of 50 gpm, or an appropriate site specific value , as an alternate Potential Loss threshold.
If used , the threshold wording should reflect that the determination of the leak rate value e1wludes normal reductions in RCS inventory (e.g., by the letdov, 1 n system or RCP seal leakoff). Potential Loss l .B Enter the site specific indications that define an e1ctreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock a transient that causes rapid RCS cooldown v1hile the RCS is in Mode 3 or higher (i.e., hot and pressurized). These 1.vill typically be parameters and values that would require operators to take prompt action to address a pressurized thermal shock condition.
Developers should also determine if the threshold needs to reflect any dependencies used as EOP transition/entry decision points or condition validation criteria (e.g., an EOP used to respond to an excessive RCS cooldown may not be entered or immediately e1dted if RCS pressure is below a certain value). For plants that have implemented Westinghouse 0 1.vners Group Emergency Response GuideliAes , enter the parameters and values used iA the RBS lH.tegrity Red Path. Because of the comple,dty of certain decision points 1 Nithin the Red Path of this CSFST , developers at these plants may elect to not include the specific parameters and values, and instead follov, 1 the guidance belov, 1* Westinghouse ERG Plants As a potential loss indication, developers should consider including a threshold the same as , or similar to, " RCS Integrity Red entry conditions met" in accordance with the guidance at the front of this section. As noted above , developers should ensure that the threshold ,vording reflects any EOP transition/entry decision points or condition validation criteria.
For e1rnmple , a threshold might read "RCS Integrity (P) Red entry conditions met with RCS pressure>
300 psig." Inadequate Heat RemeYal There is no Loss threshold associated with Inadequate Heat Removal. PWR RCS BARRIER THRESHOLDS:
144 Potential Loss 2.A l>lE I 99 0 I (Revi s ioR 6) N e vemeer 2012 This condition indicates an e J,treme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary side heat sink). This condition represents a potential loss of the RCS Barrier. If CSP H.1 was entered during a loss of coolant accident, it may be eJdted if secondary heat sink is not required based on RCS pressure less than non faulted SIG pressure. If this is the case. then declaration requirements should not be considered to be met.In accordance 1.vith EOPs , there may be unusual accident conditions during 1 Nhich operators intentionally reduce the heat removal capability of the steam generators; during these conditions , classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold 2.B; both will be met. This condition warrant s a Site Area EmergenC)' declaration because inadequate RCS heat removal may re s ult in fuel heat up sufficient to damage the cladding and increa s e RCS pressure to the point where mass 'Nill be lost from the system. CSP H.1 is the Critieal Safety Proeedure that proYides direetions if the ultimate heat sink funetion is under eJ.tref!'le ehallenge.
This eondition addresses los s of functions reEJ1:1ired for hot shl:ltdown with the reactor at pre s s1:1re and temperature and th1:1s a Potential Loss of the RCS barrier. De*;elaper Nates: Potential Loss 2.A Enter the site specific parameters and Yalues that define an e1,treme challenge to the abilit)* to remoye heat from the RCS Yia the steam generators.
These will typically be parameters and values that v r ould require operators to take prompt action to address this condition.
For plants that haYe implemented Westinghouse Owners Group Emergency Response Guidelines , enter the parameters and val1:1es used in the Heat Sink Red Path. Westinghouse ERG Plants Developers should coAsider including a threshold the same as , or similar to, " Heat SiAk Red entry condition s met" iA accordance with the guidance at the front of this se c tion. 3. RCS Activity/
Containment Radiation Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment , assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 3.A since it indicates a loss of the RCS Barrier only. There is no Potential Loss threshold associated with RCS Activity/ Containment Radiation. 145 l>ffi! 99 0 I (Re~*isieA 6) l>leverf!ser 2012 RCS BARRIER THRESHOLDS:
PWR RCS BARRIER THRESHOLDS:
Devel0t1er Nates: Loss 3.A The reading should be determiaed assumiag the iastaataneous release aad dispersal of the reactor coolaat aoele gas aad iodiae inventory , *.vith RCS activit)'
at Technical Specification allov,caele limits, iato the contaiament atmosphere.
Usiag RCS activity at Techaical Specification allowable limits aligns this threshold with IC SU3. Also , RCS activity at this Jeyel vlill t)'pically result ia containment radiation le*,cels that caa be more readily detected ey containment radiation monitors , and more readily differentiated from those caused by piping or component "shine" sources. If desired , a plant may use a lesser value of RCS activity for determi-Ring this 1 ,calue. In some eases , the site specific physical location aad sensitivity of the containmeat radiation monitor(s) may ee such that radiatioa from a cloud of released RCS gases caanot be distiaguished from radiation emanating from piping and compoaeats containing elevated reactor coolant activity.
If so, refer to the Developer
}Jotes for Loss/Potential Loss 5.A aad determiae if aR alternate indication is availaele.
: 4. Containment Integrity or Bypass 5. Not Applicable (includ e d for numb e rin g co n s i s t e n cy) Other Ietlieetiaes Loss and,lor Potential Loss 5.A This subcategory addresses other site specific thresholds that may be iacluded to indicate loss or potential loss of the RCS earrier based on plant specific desiga characteristics not considered ia the generic guidance.
14 6 RCS BARRIER THRESHOLDS:
Develeper Netes: Loss and/or Potential Loss 5.A }tE l 99 0 I (Re~*i s i e A 6) }le~*em s e r 201? Developers should determine if other reliable indicators e x ist to evaluate the status of this fi s sion product barrier (e.g., reviev,' accident analyse s de s cribed in the site Final S afety Analysi s Rep o rt , a s updated).
The goal is to identif y an y unique or site s pe c ifi c indication s th at \*lill promote timely and accurate a ss e ss ment o f barrier s tatu s. A ny added thre s hold s should represent apprmdmatel y the same relative threat to the barrier as the other thresh o ld s in this column. Basis information for the other threshold s may be u s ed to g aug e the r e lative barrier threat level. PWR RCS BARRIER THRESHOLDS:
~~Emergency Director Judgment Loss 65.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost. RCS BARRIER THRESHOLDS:
Potential Loss 5 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
De*,releper Netes: 147 L __ _ NE! 99 0 I (RevisioA
: 6) November 2012 PWR-CONTAINMENT BARRIER THRESHOLDS:
The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. _This barrier also includes the main steam, feedwater , and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
: 1. Critical Safety Function Status Tree There is no Loss threshold associated with CSFST. Potential Loss l .A This condition represents an IMMINENT core melt sequence which, if not corrected.
could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier. The restoration procedure is considered
" effective" if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.
Severe accident analyses (e.g .. NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios.
and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence. -hf.:__RCS or SG Tube Leakage Loss +2.A This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment.
The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss I.A and Loss I.A, respectively.
This condition represents a bypass of the containment barrier. FAUL TED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into , or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably
[part of the FAULTED definition]
and the faulted steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAUL TED for emergency classification purposes.
148 CONTAINMENT BARRIER THRESHOLDS:
}JEI 99 0 I (ReYisieA
: 6) }leYeml:Jer 2012 The FAULTED criterion establishes an appropr i ate lower bound on the size of a steam release that may require an emergency classification.
Steam releases of this size are readily observable with normal Control Room indications. The l ower bound for t his aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU~ for the fuel clad barrier (i.e., RCS activity values) and IC SU 4~ for the RCS barrier (i.e., RCS leak rate values). This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant , or to drive an auxiliary ( emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.
Steam releases associated with the expected operation of a SG po\Yer operated rettefatmospheric dump valve or safety relief valve do not meet the intent of this threshold.
Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown.
Steam releases associated with the unexpected operation of a valve ( e.g., a stuck-open safety valve) do meet this threshold.
149 NE! 99 01 (Re\*i s ioA 6) J>fo\*e1tt0er 2012 P\'1R CONTAINMENT BARRIER THRE8HOLD8:
Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component ( e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential Joss of containment but should be evaluated using the Recognition Category A-R I Cs. The emergency classification levels resulting from primary-to-secondary leakage , with or without a steam release from the FAULTED SG, are summar ized below. Affected SG is FA UL TED Outside of Containment?
P-to-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification (oF otheF 't'alue 13eF 8Y4 :ge't'elo13eF l-fotesj Greater than 25 gpm (oF otheF Unusual Eve nt per SU4 Unusual Eve nt per SU4 value peF 8U4 Develo13eF NetesJ Requires operation of a Site Area Emergency standby charging (makeup) per FSl Alert per FAl pump (RCS Barrier Potential Loss) Requires an automatic or Site Area Emergency Alert per FAl manual ECCS (SI) actuation per FSl (RCS Barrier Loss) There is no Potential Loss threshold associated with RCS or SG Tube Leakage. Devel0J)er N0tes: Loss 1.A ,"r steam geAerntoF 130,.veF 013ernted Fe lief valve may also be FefeFFed to as aA atmospheFic steam dump valve OF otheF apprnpFiate site specific teFm. DevelopeFS may iAclude aA additioAal site s13ecific tht=eshold(sj to addFess 13rnl0Aged steam Feleases Aecessitated by opeFatioAal coAsiderntioAs if AOPs OF EOPs could rnquirn that a leakiAg OF RUPTURE:g steam geAerntoF be used to su13p0Ft plaAt cooldovm.
:gevelo13eFs may ,vish to coAsideF iACOFJ)OrntiAg the above table iAto useF aids (e.g., a wallboaFdj OF other locatioAs withiA theiF basis documeAt.
150 PWR-CONTAINMENT BARRIER THRESHOLDS:
: 2. Inadequate Heat Remeval There is no Loss threshold associated
*with Inadequate Heat Removal. Potential Loss 2.A N E J 99 01 (Revi s ion 6) }J o v eml:Jer 2012 This condition represents an IMMINENT core melt sequence which , if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur , there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation ofa procedure(s) to restore adequate core cooling is not effective (successful) within 15 mmutes , it is assumed that the event trajectory w i ll likely lead to core melting and a subsequent challenge of the Containment Barrier. The restoration procedure is considered
" effective" if core e~dt thermocouple readings are decreasing and/or if reactor vessel level is increasing.
Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The EmergenC)' Director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effectiYe.
Severe accident analyses (e.g., l>HJREG 1150) haYe concluded that function rest o ration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is Yer)' small in these e*,ents. Given this , it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.
Develeper Netes: Some site specific EOPs and/or EOP user guidelines may establish decision making criteria concerning the number or other attributes of thermocouple readings nece s sary to drive actions (e.g., 5 CETs reading greater than 1 , 2009f is required before transit i oning to an inadequate core cooling procedure).
To maintain consistency
*with EOPs, these decision making criteria may be used in the core e~dt thermocouple reading thres h olds. Potential Loss 2.A.1 Enter site specific criteria requiring entry into a core cooling restoration procedure or prompt implementation of core cooling restoration actions. A reading of 1 , 200&deg;F on the CETs ma)' also be used. For plants that haYe implemented Westinghouse Owners Group EmergenC)'
Response Guidelines , enter the parameters and values used in the Core Cooling Red Path. 151 PWR CONTAINMENT BARRIER THRE8HOLD8:
'.Vestiaghouse ERG Plaats }IE I 99 0 I (Re\*i s i e A 6) }Je vetft be r 201 2 Developers should consider inoluding a threshold the same as , or similar to, " Core Cooling Red entry conditioas met for 15 minutes or longer" in accordance with the guidance at the front of this section. 3. RCS Activity/
Containment Radiation There is no Loss threshold associated with RCS Activity/ Containment Radiation.
Potential Loss 3 .A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the ana l ogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.
NUREG-1228 , Source Estimations During Incident Response to Sev e r e Nuclear Power Plant Accidents , indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to e*t5t;e x i st there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.
CONTiUNMENT BARRIER THRE8HOLD8:
DevelefJeF Netes: Poteatial Loss 3 .,"'L NUREG 1228 , Seur c c Estim et ien 8 DHrin g In c ident Rcspen8c te &*rcrc ]'>lucl-cer Pewcr Pkm t Accident 8, provides the basis for usin g the 20% fuel oladding failure value. Unles s there is a site specific analysis ju s tifying a different value , the reading should be determined assumiag the instaataneous release and dispersal of the reactor coolant noble gas and iodine inventory associated 1 with 20% fuel olad failure into the containmeat atmosphere.
: 4. Containment Integrity or Bypass Loss 4.A These thresholds address a situation where containment isolation is required and one of two cond ition s exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both thresholds 4.A.1 and 4.A.2. 152 PWR CONTAINMENT BARRIER THRESHOLDS:
NE T 99 0 I (ReY i s i o A 6) }>Jove A~ber 2 0 12 4.A. l -Containment integrity has been lost , i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage ( or sometimes referred to as design leakage l b).). Following the release of RCS mass into containment , containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may ( or may not) be accompanied by a noticeable drop in containment pressure.
Recognizing the inherent difficulties in determining a containment leak rate during accident conditions , it is expected that the Emergency Director will assess this threshold using judgment , and with due consideration given to current plant conditions , and available operational and radiological data (e.g., containment pressure , readi ngs on radiation monitors outside containment , operating status of containment pressure control equipment , etc.). Refer to the middle piping run of Figure 9-F-4 2_. Two simplified examples are provided.
One is leakage from a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities , the leakage could be detected by any of the four monitors depicted in the figure. 153 CONTAINMENT BARRIER THRESHOLDS:
~JE I 99 0 I (Revisio H 6) ~Jovembe r 2 01 2 Another example would be a loss or potential loss of the RCS barrier , and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of containment.
In this case , the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
Following the leakage of RCS mass into containment and a rise in containment pr e ssure , there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components.
These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category A-R ICs. 4.A.2-Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment.
As used here , the term " environment" includes the atmosphere of a room or area , outside the containment, that may, in turn , communicate with the outside-the-plant atmosphere ( e.g., through discharge of a ventilation system or atmospheric leakage).
Depending upon a variety of factors , this condition may or may not be accompanied by a noticeable drop in containment pressure.
CONTAINMENT BARRIER THRESHOLDS:
Refer to the top piping run of Figure 9-F-4 2_. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful).
There is now an UNISOLABLE pathway from the containment to the environment.
The existence of a filter is not considered in the threshold assessment.
Filters do not remove fission product noble gases. In addition , a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. P,&#xa5;R CONTAINMENT BARRIER THRESHOLDS:
Leakage between two interfacing liquid systems , by itself, does not meet this threshold.
Refer to the bottom piping run of Figure 9-F-4 2_. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building.
The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump or system piping developed a leak that allowed steam/water to enter the Auxiliary Building , then threshold 4.B would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A.1 to be met as well. 154 CONTAINMENT BARRIER THRESHOLDS:
~ffi l 99 0 I (Rev i s i e A 6) ~levem b er 20 1 2 Following the leakage of RCS mass into containment and a rise in containment pressure , there may be minor radiological releases associated with a ll owable (design) containment l eakage through various penetrations or system components.
Minor re l eases may also occur if a containment i so l ation valve(s) fails to c l ose but the containment atmosphere escapes to a closed system. These releases do n ot constitute a l oss or potential loss of containment but sho uld be evaluated using the Recognition Category A-R ICs. The status of the containment barrier during an event involving steam generator tube l eakage is assessed using Loss Thresho ld +2_.A. Loss 4.B Conta inm ent sump, temperature , pressure and/or radiation l eve l s will increase i f reactor coolant mass is leaking int o the containment.
If these parameters have not increased , then the reactor coo l ant mass may be le aking outside of containment (i.e., a containment bypass sequence).
Increases in s ump , temperature , pressure , flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings a nd a larm s on radiation monitors with detectors outside conta inment should be corroborated with other avai labl e indications to confirm that the source is a loss of RCS mass outs ide of containment.
If the fuel clad barrier has not been lost , radiation monitor r eadings outside of containment may n ot increase significantly; however , other unexpected changes in sump l eve l s, area temperatures or pressures, flow rates, etc. should be s u fficient to determine if RCS mass is being lost outside of the conta inm ent. CONTAIN 1\'IENT BARRIER THRESHOLDS:
Refer to the middle piping run of Figure 9-F-2_4. In this simp lifi ed example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locati ons and sensitivities , the leakage could be detected b y any of the four monitors depi cted in the figure and cause threshold 4.A.1 to be met as well. To ensure proper escalation of the emergency classification , the RCS leakage outside of containment must be related to t h e mass loss that is causing the RCS Loss and/or Potential Loss threshold 2_+.A to b e met. PWR CONTA~iENT BARRIER THRESHOLDS:
Potentia l Loss 4.A If containment pressure exceeds the design pressure , there exists a potential to lose the Containment Barrier. To reach this l evel , there must be an inadequate core cooling condit i on for an extended period of time; therefore, the RCS and Fuel Cla d barriers wou ld already be lost. Thus, this threshold is a discriminator between a S it e Area Emergency and Genera l Emergency since there i s now a potential to lose the third barrier. 155 CONTAINMENT BARRIER THRESHOLDS:
Potential Loss 44.B NEI 99 0 I (Revi s ioA 6) ~JOYSl'ABSF 2012 The existence of an explosive mixture means , at a minimum , that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen bum will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.
It therefore represents a potential loss of the Containment Barrier. Potential Loss 44.C This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate , and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manuall y start equipment that may not have automatically started , if possible.
This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays , ice condenser, conta i nment accident fans , etc., but not including containment venting strategies) are either lost or performing in a degraded manner. During a design basis accident a minimum of two CONTAINMENT Accident Fan Cooler Units with their accident fans running and one CONTAINMENT spray train are required to maintain the CONTAINMENT peak pressure and temperature below the design limits. Each CONTAINMENT Spray train is a CONTAINMENT spray pump, spray header. nozzles. valves and piping. Each CONTAINMENT Accident Fan Cooler Unit consists of cooling coils, accident backdraft damper. accident fan. service water outlet va l ves. and contro l s necessary to ensure an operable service water flow path. Develeper Netes: Loss 4.A.1 De11elopers may include a list of site specific radiation monitors to better define this threshold.
E1,pected monitor alarms or readings may also be included.
Potential Loss 4 .A The site specific pressure is the containment design pressure. For plants that haYe implemented Westinghouse Owners Group Emergency Response Guidelines, the pressure 11alue in Potential Loss 4 .A is that used for the Containment Red Path. If the Containment CSFST contains more than one Red Path due to other dependencies (e.g., status of containment isolat i on), enter the h ighest containment pressure value shovm on the tree. This is typica ll y the containment design pressure.
PWR CONTAINMENT BARRIER THRESHOLDS
: 156 
: 5. PoteAtial Loss 4 .B N E I 99 0 I (Re v isioA 6) ~Jo~*effieer 2012 Developers may eAter the minimum eoAtainmeAt atmospheric hydrogeA eoneentration Aeeessary to support a hydrogeA burn (i.e., the lovt'er deflagratioA limit). A eoAeurreAt eoAtaiAmeAt O>(ygeA eoAeeAtratioA may be iAeluded if the plaAt has this iAdieatioA available iA the CoAtrol Room. PoteAtial Loss 4 .C EAter the site speeifie pressure setpoiAt value that aetuates eoAtainmeAt pressure eoAtrol systems (e.g., coAtainmeAt spra)0. Also eAter the site speeific eoAtainmeAt pressure eoAtrol system/equipmeAt that should be operatiAg per design if the eoAtaiAmeAt pressure setpoiAt is reaehed. If desired , specifie eoAditioA iAdieatioAs sueh as parameter values eaA also be eAtered (e.g., a eoAtaiAmeAt spray flov,' rate less thaA a eertaiA value). This threshold is Rot applieable to the U.S. EvolutioAary Power Reaetor (EPR) desigA. WestiAghouse ERG PlaAts As a poteAtial loss iAdieation , deYelopers should eoAsider iAeluding a threshold the same as , or similar to , " CoAtainment Red eAtry coAditioAs met" iA aecordaAee 1 Nith the guidaAce at the fro At of this seetioA. Other ledieetiees Loss aAdlor PoteAtial Loss 5.A This subcategory addresses other site speeifie thresholds that may be iAeluded to iAdieate loss or poteAtial loss of the ContairuneAt barrier based oA plaAt specific design eharacteristics not considered iA the geAeric guidaAee.
CONTAINMENT BARRIER THRESHOLDS:
DeYelefJer Netes: Loss and/or PoteAtial Loss 5 .A If site emergeAC)' operatiAg proeedures provide for veAtiAg of the eoAtaiAmeAt as a meaAs of pre*,'eAting catastrophic failure , a Loss threshold should be included for the eoAtaiAmeAt barrier. This threshold would be met as sooA as such veAting is IMMl1'ffiNT. Containment veAtiAg as part of reeovery actions is elassified in aeeordanee with the radiologieal efflueAt ICs. Developers should determiAe if other reliable iAdicators e1dst to evaluate the status of this fission product barrier (e.g., review aecideAt aAalyses deseribed iA the site FiAal Safety AAalysis Report , as updated).
The goal is to ideAtify any uAique or site specifie iAdieatioAs that 1+'l'ill promote timely aAd aecurate assessmeAt of barrier status. PWR CONTAINMENT BARRIER THRESHOLDS:
157 NE I 99 0 I (R ev i s i o A 6) ~love 1T10 e r 2012 Any added thre s h o ld s should represent approximately the same relative threat to the barrier as the other threshold s in this column. Basis information for the other thresholds may be used to g au g e the relative barrier threat level. &~Emergency Director Judgment Loss 65.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost. Potential Loss 65 .A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
DeYelef)er Note s: 158 Inside Containment Damper RCP Seal Cooling Figure 9-F-3 24: P-WR-Containm e nt Integri ty or B y pas s Exa mpl es Auxiliary B uil ding I : : Effluent , Damper Penetration J' t ' I Ope n va l ve : Process : I I : Monitor : 159 Closed Cooling Water System }lei 99 01 (RevisieA 6) }leYeFAeer 20 J?
N e l 99 0 I (Revision
: 6) Oetoi:Jer 2011 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS Table H 1: Ree0gniti0n Categery "H" Initiating C0nditi0n Matrix UNUSUAL EVENT HUl Confirmed SECURITY CONDITION or threat. Op. },fades: All HU2 Seismic event greater than OBE levels. Op. },fades: All HU3 Hazardous event. Op. },fades: All HU4 FIRE potentially degrading the le*,rel of safety of the plant. Op. },fades: All HU7 Other conditions e,cist which in the judgment of the Emergency Director warrant declaration of a (NO)UE. Op. },fades: All ALERT W..J HOSTILE ACTION *n<ithin the OWNER CONTROLLED AREA or airborne attack threat 1.vithin 30 mim1tes. Op. },{edes: All HA5 Gaseous release impeding access to equipment necessary for noffilal plant operations , eooldown or shutdown.
Op. Uedes: All HA6 Control Room evacuation resulting in transfer of plant contro l to alternate locations. Op. }.{edes: All W ... 7 Other conditions exist which in the judgment of the Emergency Director warrant declaration ofan Alert; Op. },fades: All SITE}* ... REA EMERGENCY HS1 HOSTILE ACTION v,cithin the PROTECTED AREA. Op. },fedes: All GENERAL EMERGENCY HGl HOSTILE ACTION resulting in loss of physical control of the facility.
Op. },fades: All ,-------------------, : Table intenaea for use b:Y I EAL ae:;reJ013ers.
: Inclusion in licensee I ,.I * * ,.I , uocul'l'l.ents ts not requ1reu. L------------------J HS6 Inability to control a key safety flmction from outside the Control Room. Op. },{edes: All HS7 Other conditions e)dst \Vhich in the judgment of the Emergency Director warrant declaration of a Site Area Emergency. Op. },fades: All 160 HG7 Other conditions eJdst which in the judgment of the Emergency Director warrant declaration of a General Emergency.
Op. },fedcs: All ECL: Notification of Unusual EventUnusua l Event Initiating Condition:
Confirmed SECURITY CONDITION or threat. Operating Mode Applicability:
All Emergency Action Levels: }I E I 99 01 (RevisieA 6) }leveffieer 2012 HU1 EJrnmple Emergency Action Levels: (l Of 2 Of 3) A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the (site specific s.S.ecurity sh-tft-Shift supervisionSupervisor1. Notification of a credible security threat directed at the sitePBNP. A validated notification from the NRC providing information of an aircraft threat. Definitions:
SECURITY CONDITION:
Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel.
or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION. HOSTILE ACTION: An act toward PBNP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. Th i s includes attack by air, land, or water using guns. exp l osives, PROJECTILEs.
vehicles.
or other devices used to deliver destructive force. Other acts that satisfy the overa ll intent may be i ncluded. HOSTILE ACTION should not be construed to include acts of civil disobedience or fe l onious acts that are not part of a concerted attack on the nuclear power plant. -terrorism-based EALs should be used to address such activities (i.e., this may inc l ude violent acts between ind i viduals in the owner controlled area). SAFETY SYSTEM: A system required for safe plant operation, cool i ng down the plant and/or p l acing it in the cold shutdown condit i on. i ncluding the ECCS. These systems are classified as safety-related. Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR -&sect;--73. 71 or 10_-CFR~ 50.72. Security events assessed as HOSTILE ACTIONS are classifiab l e under ICs HAI , HSl and HGl. Timely and accurate communications between Security Shift Superv i sion Supervisor and the Control Room i s essential for proper classification of a security-related event. Classification of 161 l>lEI 99 0 I (Re>risieA
: 6) Nm*ember 20 1 2 these events will initiate appropriate threat-related notifications to plant personnel and GRGoffsite response organization
: s. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan , Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage In stallation Security Program}.
162 
~I E I 99 01 (RevisieA 6) ~le veffieer '.lQ 12 EAL HU 1.1 references (site specific s,S.ecurity shlft-Shift supervisionSupervisor1 because these are the individuals trained to confirm that a security event is occurring or has occurred.
Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR &sect; 2.39 Q information. EAL HUI.2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with SY-AA-I 02-1014. Threat Assessment and Reporting(site specific procedure). EAL HU I .3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.
The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with (site specific procedure)AOP-29.
Security Threat. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HAI. Developer Notes: The (site specific secHrity shift supervision) is the title of the on shift individual respons i ble for supervision of the on shift security force. The (site specific procedure) is the procedure(s) used by Control Room. and/or Securit)* personnel to determine if a security threat is credible , and to validate receipt of aircraft threat information.
Emergency plans and implementing procedures are pHblic documents; therefore , EALs should not incorporate Security sensitive information.
This i0cludes information that may be adva0tageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
Securit)* sensitive i0formation should be contained in non public documents such as the Securit)* PlaA. With due consideration given to the above developer note , EALs may contain alpha or numbered references to selected events described in the Security Plan and associated implementing procedures. Such references should not contain a recognizable description of the e*,rent. For e1rnmple , an EAL may be *.vorded as " Security event #2 , #5 or #9 is reported by the (site specific security shift supervision)." EGL Assignment Attributes:
3 .1.1.A 163 l>J e l 99 0 I (RevisieA
: 6) J>leYeffiser 2012 ECL: Notification of Unusual Event Unusua l Event Initiating Condition:
Seis mic event grea ter than OBE level s. Operating Mode Applicability:
All Example Emergency Action Levels: HU2 H 2.1 Seismic event greater than Operating Ba sis Ea rthquak e (OBE) as indicated by seismic monitor indication of ground acceleration greater than: * -0.06 g horizontal
* 0.04 g vertical.
(1) (site specific inelication that a seismic event met or e1weeeleel OBE limits) Definitions:
Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OB E) 1* An earthquake greater than an OBE but le ss than a Safe Shutdown Earthquake (SSE)2 s hould have no significant imp act on related systems, structures and components; however , some tim e may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., perform s wa lk-downs and post-ev ent inspections
). Given the time necessar y to perform walk-do wns and inspections , and fully under s tand any impacts , this event represents a potential degrad a tion of th e l ev el of safe ty of the plant. Event verification with ex ternal sources shou ld not be neces sa r y during or fo llowing an OB E. E arthquakes of this magnitude should b e readily felt b y on-site personnel and recogni ze d as a seismic event (e.g., t:ypical lateral accelerations are in e1wess of 0.08g). T he Shift Manager or Emergency Director may seek external verification if deemed appropriate ( e.g., a call to the USGS, check internet news sources, etc.); how ever, the verific a tion action must not preclude a timely emergency declaration.
1 A n OB E is vi bratory ground m otio n fo r whic h those features of a nucl ear power plant n ecessa r y fo r co ntinu e d opera ti o n w ith o u t undue r i sk to the h ea l t h a nd safety of th e pub lic w ill r e m a in functional.
2 A n SSE is v ib rato r y gro und motion for w h ic h ce rt a in (ge n e r a ll y, safety-r elate d) structures , syste m s, a nd co mpon e nt s must b e d es i g n ed t o r e main fun c tional. 164 
~l e i 99 01 (Re,*isieR 6) ~10V6FR061'
: 20) 2 Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9. DeYelaper Nates: This "site specific iAdicatioA that a seismic eveAt met or eJweeded OBE limits" should be based OR the iAdicatioAs , alarms aAd displays of site specific seismic moAitoriAg equipmeAt.
lAdicatioAs described iA the EAL should be limited to those that are immediate!)
' a*railable to CoAtrol Room persoAAel aAd ',&#xa5;hich caA be readily assessed. lAdicatioAs available outside the CoAtrol Room aAd/or which require leAgthy times to assess (e.g., processiAg of scratch plates or recorded data) should not be used. The goal is to specify indicatioAs that caA be assessed within 15 miAutes of the actual or suspected seismic eveAt. For sites that do not hav e readily assessable OBE iAdicatioAs vrithiA the Control Room , developers should use the followiAg alternate EAL (or similar wordiAg).
(1) a. b. CoAtrol Room persoAAel feel aA actual or poteAtial seismic e1,*eAt. The oc c urreAce of a seismic eYent is confirmed iA manAer deemed appropriate by the Shift Manager or EmergeAey Director.
The EAL 1.b statement is included to ensure that a deelaratioA does Rot result from felt YibratioAs caused by a noA seismic source (e.g., a dropped heavy load). The Shift MaAager or EmergeAC)' Director may seek eJC:ternal YerificatioA if deemed appropriate (e.g., a call to the USGS , check internet Aews sources, etc.); howe .. *er , the verification actioA must Rot preclude a timely emergency declaration.
It is recogAized that this alternate EAL wordiAg may cause a site to declare an Unusual E YeAt 1 , 1 ,*hile aAother site , similarly affected but with readily assessable OBE indicatioAs ie the CoAtrol Room , may not. The above alternat e wording may also be used to develop a compeAsatory EAL for use during periods whee a seismic moAitoring system capable of detecting aA OBE i s out of service for maiAteAance or repair. EGL AssigAmeAt Attributes:
3 .1.1.A 165 
}l e i 99 0 I (R e vi s i e A 6) }l0\'0FABeF 20} 2 HU3 ECL: Notification of Unusual EventUnusual Ev ent Initiating Condition:
Hazardous event~ Operating Mode Applicability:
All Emergency Action Levels: Example Emergene)'
A.eti0n LeYels: (1 or 2 or 3 or 4 or 5 or 6) Note: EAL HU3.~ does not apply to routine traffic impediments such as fog , snow , ice , or vehicle breakdowns or accidents. ~1 ~2 H R 3.3 A tornado strike within the PROTECTED AREA. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode. Movement of personnel within the PROTECTED AREA is impeded due to an off site event involving hazardous materials ( e.g., an off site chemical spill or toxic gas release).
A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles. (Site specific list of natural or technological hazard events)Lake level greater than or equ a l to +8.0 ft. (Plant elevation).
Pump bay level le s s than -15 .0 ft. Definitions:
PROTECT E D AREA: The area under continuou s acces s monitoring and control, and armed protection as described in the site Security Plan. SAFETY SYSTEM: A system required for s afe plant operation. cooling down the p l ant and/or placing it in the cold shutdown condition, including the E CCS. These systems are classified as safety-related. Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL HU 3.l addresses a tornado striking (touching down) within the Protected Area. EAL HU3 .2 addresses flooding of a building room or area that results in operators iso l ating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation 166 J>le l 99 O I (Re~*isieA 6) J>le v e mber ?Q 12 of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification , operability of the affected component must be required by Technical Specifications for the current operating mode. EAL HU3 .3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA. EAL HU3.4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles.
Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure , etc., or an on-site train derailment blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow , ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992 , the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. EAL HU3.5 addresses lake water level as the Turbine Building is susceptible to external flooding.
+8.0 ft.correspondsft.
corresponds to the Turbine Building floor elevation(site specific description). EAL HU3.6 addresses operability of the Emergency Die s el Generators and all Containment fan coolers. The pump bay level of -15.0 ft.representsft. repre se nts the value at which the Emergency Diesel Generators and all Containment Fan Cooler s must b e declared inoperable, and is four feet above the level at which the service water pumps may begin to cavitate.
Escalation of the emergency classification level would be based on I Cs in Recognition Categories AR , F, S or C. Develeper Netes: The " Site specific list of natural or technological hazard events" should include other events that may be a precursor to a more significant event or condition , and that are appropriate to the site location and characteristies. Notwithstanding the events specifically included as EALs above , a " Site specific list of natural or technological hazard events" need not include short lived events for which the eJ{tent of the damage and the resulting consequences can be determined within a relatively short time frame. In these cases , a damage assessment can be performed soon after the event , and the plant staff *will be able to identif)' potential or actual impacts to plant systems and structures.
This will enable prompt definition and implementation of compensator y or corrective measures *w i th no appreciable increase in risk to the public. To the eJ{tent that a short lived event does cause immediate and significant damage to plant systems and structures , it v 1 ill be classifiable under the Recognition Category F , S and C ICs and EALs. Events oflesser impact would be e1{pected to eause only small and localized damage. The consequences from these types of events are adequately assessed and addressed in accordance with Technical Specifications.
In addition , the occurrence or effects of the event may be reportable under the requirements of 10 CFR 50.72. EGL Assignment Attributes:
3.1.1.A and 3.1.l.C 167 T>JEI 99 01 (RevisioA
: 6) November 2012 HU4 ECL: Notification of Unusual EventUnusual Event Initiating Condition:
FIRE potentially degrading the level of safety of the plant. Operating Mode Applicability:
All Emergency Action Levels: Example Emergeeey
," .. etiee Levels: (1 or 2 or 3 or 4) Note~: H LJ 4.1 H 4.2 H 4.4 _* _The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time ha s been exceeded , or will likely be exceeded.
* A Containment fire alarm is considered valid upon receipt of multiple zones (more than 1) on the F ACP system (this note is applicable in Modes 1 and 2 only). a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications
:
* Report from the field (i.e., visual observation)
* Receipt of multiple (more than 1) fire alarms or indications
* Field verifica tion of a single fire alarm AND b. The FIRE is located within ANY of the follov,ingTable H-1 plant rooms or areas-:-a. (site specific list of plant rooms or areas) Receipt of a single fire alarm ~with no other indications of a FIRE). ----AND b. The FIRE is located within ANY of the followingTable H-1 plant rooms or areas except Containment in Modes 1 and 2 (see Note above): (site specific list of plant rooms or areas) ----AND c. The existence of a FIRE is not verified within 30-mi nutes of alarm receipt. A FIRE within the plant or ISFSI [ferpl-cmts with 819 ISFS! eutsidc the pf.cm! Preteeted Aff8}-PROTECTED AREA not extinguished within 60-_: minutes of the initial report, alarm or indication. A FIRE within the plant or ISFSI [ferpkmts with 8n ISFSI eutside the pt8nt Pmteeted Aff8}-PROTECT E D AREA that requires firefighting suppo rt by an offsite fire response agency to extinguish.
Table H-1 Areas Control Room Containment PAB 168 GOS building I 3.8kV Building Cab l e Snreading Room Vital Switchgear Room AFW Pumn Room G-0 1/02 Rooms EDG Building Service Water Pumn Rooms Fa&#xa3;ade 85' 169 }le! 99 0 I (ReYisieA
: 6) }leveffiaer 2012 Definitions:
1>J EI 99 01 (ReYi s ioR 6) 1>J ovember 2012 FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. PROTECTED AREA: The area under continuous access monitoring and control. and armed protection as described in the site Security Plan. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a pot e nti al degradati on of the level of safe ty of the plant. With regard to containment fire alarms. there is constant air movement in containment due to the operation of the air handling system drawing air to the cooling units past the smoke detectors. It can reasonably be expected that a fire that burns for 15 minutes would produce sufficient products of combustion to cause fire detectors in multiple zones to alarm. EAL HU4.l The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper ba ske t). In addition to alarm s, other indications of a FIRE could be a drop in fire main pressure , automatic activation of a suppression system, etc. Upon receipt , operators will take prompt actions to confirm the validity of an initial fire alarm, indication , or report. For EAL assessment purpose s, the emergency declaration clock starts at the time that the initial alarm, indication , or report was received , and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time ofreceipt of the initial alarm, indication or report. EAL HU4.2 This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., prov e d or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes , the 3 0-minute clock starts at the time that the initial alarm was recei ve d , and not the time th at a subsequent ver ification action was performed.
A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason , additional time is allowed to verify the validity of the alarm. Except for the Containment Building in Modes 1 or 2. t+he 30minute period is a reasonable amount of time to determine if an actual FIRE exists; however , after that time, and absent information to the contrary, it is assumed that an actual FIRE i s in progress. 170 l>l e l 99 O I (Re,*isieR 6) l>l evember 2012 If an actual FIRE is verified by a report from the field , then EAL HU4.1 is immediately applicable , and the emergency must be declared if the FIRE is not extinguished within 15.:minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation , and this verification occurs within 30-minutes of the receipt of the alarm , then this EAL is not applicable and no emergency declaration is warranted.
EAL HU4.3 In addition to a FIRE addressed by EAL HU4.1 or EAL HU4.2 , a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of att the ISFSI located outside the plant PROTECTED AREA. [Sentence for plants *with an ISFSI o u tside the plant Protected Area] EAL HU4.4 If a FIRE within the plant or ISFSI [forpkmts with en !SFSI eutside theplemt Protected Arce] PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency ( e.g., a local town Fire Department), then the level of plant safety is potentially degraded.
The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish.
Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Basis-Related Requirements from Appendix Rand NFPA-805 Criterion 3 of Appendix A to 10 CFR 50 states in part that "structures, systems. and components important to safety sha ll be designed and located to minimize.
consistent with other safety requirements.
the probabi l ity and effect of fires and explosions." The Nuclear Safety Goal (''NSG") i n NFPA 805, Section 1.3.1 states, " The nuclear safety goal is to provide reasonable assurance that a fire during any operat i onal mode and plant configurat i on will not prevent the pla n t from achieving and ma i ntaining the fue l in a safe and stable condition." When considering the effects of fire. those systems associated with achieving and maintaining safe shutdown conditions assume ma j or importance because a safe shutdown success pat h , free of fire damage. must be available to meet the nuclear safety goals. objectives and performance criteria for a fire under any plant operational mode or configuration.
Because fire may affect safe shutdown systems and because the loss of funct i on of systems used to mit i gate the consequences of design basis accidents u n der post-fire conditions does not per se i mpact pub l ic safety. the need to limit fire damage to systems requ i red to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.
171 
}le i 99 O I (R e~*i s i o n 6) }lo v e ffieer 20 I 2 In addition, Appendix R to 10 CFR 50. requires, among other considerations, the use of I-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). Even though PBNP has adopted the alternate approach provided by NFPA-805 in lieu of the deterministic requirements of Appendix R. the 30-minutes to verify a single alarm as used in EAL HU4.2 is considered a reasonab l e amount of time to determine if an actual FIRE ex i sts without presenting a challenge to the nuclear safety performance criteria.Basis Related Requirements from AppendiJ, R AppendiJ, R to 10 CFR 50 , states in part: Criterion 3 of Appendix.
A to this part specifies that "Structures , systems , and components important to safety shall be designed and located to minimize , consistent with other safety requirements , the probability and effect of fires and eJ,plosions." When considering the effects of fire, those systems associated with achieving and maintain i ng safe shutdown condit i ons assume major importance to safety because damage to them can lead to core damage resulting from loss of coo la-Rt through boil off. Because fire may affect safe shutdown systems a-Rd because the loss of function of systems used to mitigate the consequences of design basis accidents under post fire conditions does not per se impact public safety , the need to l imit fire damage to systems required to achie*,e and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems requi-red to mitigate the consequences of design basis accidents.
In addition, Appendbt R to IO CFR 50 , requires, among other considerations , the use of 1 hour fire barriers for the enclosure of cable and equiprAent and associated non safety circuits of one redundant train (G.2.c). As used in EAL #2 , the 30 minutes to verify a single alarm is 1.vell within this worst ease 1 hour time period. Depending upon the plant mode at the time of the event , escalation of the emergency classification level would be via IC CA6 or SA9. DeYeleper Notes: The " site specific list of plant rooms or areas" should specify those rooms or areas that contain SA.,Ii'ETY SYSTEM equipment.
As noted in the EALs and Basis section , include the term ISFSI if the site has an ISFSI outside the pla-Rt Protected Area. EGL Assignment Attributes:
3.1.1.A 172 ECL: Notifica-tion of Unt1sual EventUnusual E v e nt J>Je l 99 0 1 (R e vi s i e A 6) J>Jeye ffleeF 2 0 12 HU7 Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a fNGj UE. Operating Mode Applicability:
All l E,e1Hple Emergency Action Leve l s: 1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive materia l requiring offsite response or monitoring are expected unless further degradation of SAF E TY SY S T E MS safety S)'Stems occurs. Definitions:
SAFETY SYSTEM: A system required for s afe plant operation.
cooling down the plant and/or placing it in the cold shutdown condition, including the E CCS. These systems are classified as safety-related.
Basis: This IC addresses unanticipated conditions not addressed explicitly e l sewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a tlOUEU E. 173 ECL: Alert J>J el 99 O I (R e\*i s i e A 6) J>Je YeFAe e r 2012 HA1 Initiating Condition:
HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. Operating Mode Applicability:
All Examf)le Emergency Action Levels.:.: (1 or 2) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the (site specific s.S.ecurity 5-htft-Shift supervisionSupervisor). A validated notification from NRC of an aircraft attack threat within 30 minutes of the site. Definitions:
HOST AGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION: An act toward PBNP or its personnel that includes the use of violent force to destroy equipment.
take HOST AGES, and/or intimidate the licensee to achieve an end. This includes attack by air. land. or water using guns, e x plosives.
PROJECTILES. vehicles.
or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e .. this may include violent acts between individuals in the owner controlled area). OWNER CONTROLL E D AREA: The site property owned by or otherwise under the control of the licensee.
PROJECTILE:
An object directed toward a nuclear power plant that could cause concern for its continued operability.
reliability.
or personnel safety. Basis: HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTIO:N: An act tov,ard PRNP or its personnel that includes the use of violent force to destroy equipment.
take HOSTAGES, and/or intimidate the licensee to achie,*e an end. This includes attack by air. land. or water using guns. e1cplosives.
PROJECTILE, vehicles.
or other devices used to deliver destructive force. Other acts that satisf)* the overall intent may be included.
HOSTILE ACTIO}>~ should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPPnuclear po*Ner plant. Non terrorism based EALs should be used to address such activities (i.e .. this may include violent acts between individuals in the owner controlled area). 174 
:t>le l 9 9 01 (RevisioA 6) :t>l o\'eA'leer 2012 0'.Vl>rnR COl>ffROLLED AREA: The site property owned by, or otherwise under the control of. the licensee. PROJECTILE:
An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA , or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications between Security Shift Supervision Supervisor an d the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12 , Template for the Security Plan, Training and Qualification Plan , Safeguards Contingency Plan [and Independ e nt Spent Fu e l Storage Installation Security Program}.
As time and conditions allow , these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures ( e.g., evacuation , dispersal or sheltering).
The Alert declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions. This IC does not apply to incidents that are accidental events , acts of civil disobedience , or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft , shots from hunters , physical disputes between employees , etc. Reporting of these types of events is adequately addressed by other EALs , or the requirements of 10 CFR&sect;-73.71 or 10 CFR&sect;-50.72. EAL HAI.I is applicable for any HOSTILE ACTION occurring , or that has occurred , in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA. EAL HA 1.2 addresses the threat from the impact of an aircraft on the plant , and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and GRGoffsite response organization s are in a heightened state of readiness.
This EAL is met when the threat-related information has been validated in accordance with fSY-AA-102-1014.
Threat Assessment and Reportingsite specific procedure).~
The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.
The status and size of the plane may be provided by NORAD through the NRC. In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected , although not certain , that notification by an appropriate Federal agency to the site would clarify this point. In this case , the appropriate federal agency is intended to be NORAD , FBI, FAA or NRC. The emergency declaration , including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency. 175 N e l 99 O I (Re~<isieA 6) ~Je~*emser ?Q 12 Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This inc l udes information that may be advantageous to a potential adversary , such as the particu l ars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HS 1. Develeper Netes: The (site specific sect1rity shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security sensitive information.
This includes information that may be ad11aAtageous to a potential ad11ersary , such as the particulars concerning a specific threat or threat location.
Security sensiti11e information should be contained in non public documents such as the Security Plan. '>lith due consideration gh 1 en to the abo11e de*,eloper note, EALs may contain alpha or numbered references to selected e11ents described in the Security Plan and associated implementing procedures.
Such references saould not contain a recognizable description of the e11ent. For e:itample , an EAL may be 1 Norded as "Security event #2, #5 or #9 is reported by the (site specific securit)' shift supert'ision)
." See the related De*,eloper Note in Appendix B, Definitions , for guidance on the de11elopment of a scheme definition for the O\l/NER CONTROLLED AReA. EGL Assignment Attributes:
3.1.2.D 176 l>JE I 99 01 (Revi s i e H 6) l>Jeyemeer 2012 HAS ECL: Alert Initiating Condition:
Gaseous release impeding access to equipment necessary for normal plant operations , cooldown , or shutdown.
Operating Mode Applicability:
All Example Emergency Act ion Levels: Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred , then no emergency classification is warranted.
H~5.l a. Release of a toxic , corrosive, asphyxiant or flammable gas into any of the following plant rooms or areas: Area Mode Ul VCT Area 3/4/5 U2 VCT Area 3 /4 I 5 UI Primarv Samnle area 3 U2 Primarv Samole area 3 CCWHXRoom 4/5 C-59 area 3/4/5 Pioewav 2. 8 ft. Elev. 3/4 Pinewav 3. 8 ft. Elev. 3/4 1/2B32 MCC Area 4 ---_(site specific list of plaAt rooms or areas with eAtry related mode applicability ideAtified)
AND b. Entry into the room or area is prohibited or impeded. Definitions:
Basis: This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation , or required for a normal plant cooldown and shutdown.
This condition represents an actual or potential substantial degradation of the level of safety of the plant. An Alert declaration is warranted if entry into the affected room/area is , or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release. 177 T>l~I 99 0 I (R e~*i s i e A 6) T>leveff!b er 201 2 Evaluation of the IC and EAL do not require atmospheric sampling; it on l y requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly imp ede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard ana l ysis, report of ill effects on personnel , advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs , that i s not routinely employed).
An emergency declaration is not warranted if any of the following conditions apply.
* The plant is in an operat in g mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release).
For example , the plant is in Mode 1 when the gaseous release occurs , and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
* The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area ( e.g., fire suppression system testing).
* The action for which room/area entry is required is of an administrative or record keeping nature ( e.g., normal rounds or routine inspections).
* The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action. An asphyxiant i s a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment.
_This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness , or even death. The list of plant room s or areas in EAL HA5.1 w as generated from a s t e p-b y-step revi ew of OP-3A, 3B. 3C, 5D, and 7 A. This EAL does Rot appl y to firefightiAg activities that automaticall y or maAually activate a fire suppres s ioA s y s tem iA aA area , or to iAteAtioAal iAertiAg o f coAtaiAmeA t (BWR ORiy). Escalation of the emergency classification level would be via Recognition Category AR , C or F ICs. This list wa s geAerated fr om a step by s tep review of OP 3A. B. aAd C. Develaper Nates: The " site specific list of plaAt rooms or areas with eAtry related mode applicability ideAtified" should specify those ro o m s or areas that c o AtaiA equipmeAt which require a maAual/local actioA as specified iA operatiA g procedures u s ed for Aormal plaAt operatioA , coold o wA aAd shutdowA.
Do Rot iAclude rooms or area s iA which actioAs of a coAtiAgeAt or emergeAc y Aature *would be performed (e.g., aA actioA to addres s aA off Aormal or emergeAcy coAditioA such as emergeAcy repairs , corrective measure s or emergeAcy operatioAs).
IA additioA , the list should specify the plaAt mode(s) duriAg which eAtry would be required for each room or area. 178 J>le l 99 01 (RevisieA 6) J>le veffiser 2012 The list should not include rooms or areas for 1 Nhich entry is required sole!)' to perform actions ofcm administrative or record keeping nature (e.g., normal rounds or routine inspection s). The list need not include the Control Room if adequate engineered safet),ldesign features are in place to preclude a Control Room eYacuation due to the reiease of a ha2:ardous gas. Such features may include , but are not limited to , capability to draw air from multiple air intakes at different and separate locations , inner and outer atmospheric boundaries , or the capability to acquire and maintain p os itive pre s sure within the Control Room enYelope.
If the equipment in the li s ted room or area was already inoperable , or out of service , before the event occurred , then no emergency should be declared since the event will have no adverse impact be y ond that alread y allov,'ed by Technical Specifications at the time of the event. EGL Assignment Attributes:
3 .1.2.B 179 180 l>J61 99 0 I (Re~*isieR 6) J>Je~*eFReer 2012 ECL: Alert ~/J;I 99 0 I (Re\*isieA 6) ~le Yeml:Jer 2012 HA6 Initiating Condition:
Control Room evacuation resulting in transfer of plant control to alternate locations. Operating Mode Applicability:
All Example Emergency Action Levels: H 6.1 An event has resulted in plant control being transferred from the Contro l Room to specific remote shutdovm panels and local control stations)AOP local control stations. Definitions:
Basis: This IC addresses an evacuation of the Control Room that results in transfer o f plant con t rol to alternate locations outside the Control Room. The l oss of the ability to control the plant from the Control Room i s considered to be a potential substant i a l degradation in the level of plant safety. Following a Control Room evacuation , control of the plant will be transferred to a lt ernate shutdown locations.
The necessity to control a plant shutdown from outside the Control Room , in addition to responding to the event that required the evacuation of the Control Room , will present challenges to plant operators and other on-sh ift personnel.
Activation of the ERO and emergency response facilities will assist in responding to these challenges.
Escalation of the emergency classification level would be via IC HS6. Devel0peF Notes: The "site specific remote shutdovm panels and local control stations" are the panels and control stations referenced in plant procedures used to cooldovm and shutdown the plant from a location(s) outside the Control Room. EGL Assignment Attributes:
3.1.2.B 181 ECL: Alert N f.l 99 01 (Re~*isieA 6) J>JeYeffieer 2012 HA7 Initiating Condition:
Other conditions exist which in the judgment of the E mergency Director warrant declaration of an Alert. Operating Mode Applicability:
All l E,emple EmergeeeyEmergency Action Levels: 1 Other conditions exist which , in the judgment of the Emergency Director , indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guide line exposure levels. Definitions:
HOSTILE ACTION: An act toward PBNP or i ts personnel that includes the use of violent force to destroy equipment take HOSTAGES, and/or intimidate the licensee to achieve a n end. T hi s includes attack by air. land. or water u sing guns. explosives.
PROJECTILEs.
vehicles.
or ot h er devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTrON should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. terrorism-based EALs should be used to address such activities (i.e., th i s may inc l ude violent acts between individuals in the owner contro ll ed area). Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because cond ition s exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert. 182 ECL: Site Area Emergency Initiating Condition:
HOSTILE ACTION within the PROTECTED AREA. Operating Mode Applicability:
All Examf)le EmergeneyEmergencv Action Levels: ~m l 99 01 (ReYisieA
: 6) Nevernber 2012 HS1 I S l.1 A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site specific s,S.ecurity sh+f:t-Shift supervisionSupervisor:,. Definitions:
HOSTILE ACTION: An act toward PBNP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air. land, or water using guns. explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power p l ant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals i n the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception.
equipped with suitable weapons capable of killing, maiming. or causing destruction.
INDEPENDENT SPENT FUEL STORAGE TNSTALLA TION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. PROJECTILE:
An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability.
or personnel safety. PROTECTED AREA: The area under cont i nuous access monitoring and control. and armed protection as described in the site Security Plan. Basis: This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.
Timely and accurate communications between Security Shift &#xa3;1:1pervision Supervisor and the Control Room is essential for proper classification of a security-related event. 183 T>IE: I 99 0 I (ReYi s i e A 6) T>l eveffiser 2012 Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan , Training and Qualification Plan , Safeguards Contingency Plan [and Independ ent Spent Fuel Storage Installation Security Program}.
As time and cond iti ons al l ow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).
The S i te Area E mergency declaration w ill mobilize GR:Goffsite response organization re so urces and have them available to develop and implement pub li c protective actions in the unlikel y event that the attack is successful in impairing multiple safety functions.
Th i s IC does not app l y to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA l ocated outside the plant PROTECTED AREA; such an attack shou ld be assessed using IC HAI. It a l so does not app l y to incidents that are accidental events, acts of civil disobedience , or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examp l es include the crash of a sma ll aircraft, shots from hunters , physical disputes between employees, etc. Reporting of these types of events is adequate l y addressed by other EALs , or the requirements of 10 CFR-&sect;-73.71 or 10 CFR-&sect;-50.72. Emergency plans and implementing procedures are pub li c documents; therefore, EALs should not incorporate Security-sensit i ve information.
This includes information that may be advantageous to a potential a dv ersary, suc h as the particulars concern ing a specific threat or threat lo cation. Security-sensitive in format i on s h ou ld be contained in non-public do c um ents suc h as the Security Plan. Esca l ation of the emergency classification l evel would be via IC HG 1. De>;elepeF Netes: The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Securit)'
sensitive in.formation.
This inc l udes information that may be advantageous to a potential adversary , such a s the particulars concerning a specific threat or threat location.
Security sensitive information should be contained in non public documents such as the Security Plan. With due consideration given to the above developer note , EALs may contain alpha or numbered references to selected events described in the Security Plan and associated implement i ng procedures.
Such references should not contain a recognizable description of the event. For example , an EAL may be .,.,,orded as " Security event #2 , #5 or #9 is reported by the (site specific security shift supervision)." See the related Developer Note in Appendi>L B , Definitions , for guidance on the development of a scheme definition for the PROTECTED AREA. EGL Assignment Attributes:
3.1.3.D 184 ECL: Site Area Emergency l>H s l 99 0 I (Revi s i o R 6) l>io vember ?Q 12 HS6 Initiating Condition:
Inability to control a key safety function from outside the Contro l Room. Operating Mode Applicability:
All Examf)le Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that (site specific number ofl 5 minutes 1 has been exceeded, or will likely be exceeded. a. An event has resulted in plant control being transferred from the Control Room to (site specific remote shutdown panels andAOP local control stations 1. AND b. Control of ANY of the fo ll owing key safety functions is not reestablished within (site specific number ofl5 minutes 1.
* Reactivity control
* Core cooling [PWR] I RPV 1.vater leYel [BWR]
* RCS heat removal Definitions:
Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to ga in control of a key safety function following a transfer of plant contro l to alternate locations i s a precursor to a challenge to one or more fission product barriers with in a relativel y short period of time. The determination of whether or not " control" is estab li shed at the remote safe shutdown location(s) is based on E mergency Director judgment.
_The Emerge nc y Director is e xpec ted to make a reasonable , informed judgment within (the site specific time for transfer).12 minutes whether or not the operating staff ha s control of key safety functions from the remote safe shutdown location(s).
Escalation of the emer ge ncy classification l evel wou ld be via IC FGl or CGl. DeYelef)er Notes: The " site specific remote shutdown panels and local control stations" are the panels and control stations referenced in plant procedures used to coo l down and shutdown the plant from a l ocation(s) outs i de the Control Room. 185 1>JE I 99 0 1 (R e Yi s i e fl 6) 1>Jove mber 2012 The " site specific number of minutes" is the time in which plant control must be (or is e~(pected to be) ree s tablished at an alternate location as de s cribed in the site specific fire response analyse s. Ab s ent a basi s in the site spe c ifi c anal y s es , 15 minutes should be u s ed. Another time period n1ay be used with appropriate basi s/ju s tification. EGL AssignmeAt Attributes:
3.1.3.B 186 ECL: Site Area Emer g ency N E I 99 0 I (ReYisieR 6) l>le veffiber 2012 HS7 Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.
Operating Mode Applicability:
All / Example Emergency Action Levels: WL I Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts , (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which e x ceed EPA Protective Action Guideline exposure levels beyond the site boundary.
Definitions:
HOSTILE ACTION: An act toward PBNP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air. land, or water using guns. explosives.
PROJECTIL E s. vehicles, or other devices used to deliver de s tructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. terrorism-based EALs should be used to address such activities (i.e .. this may include violent acts between individual s in the owner controlled are a). Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a Site Area Emergency. 187 
}I E I 99 0 I (Revi s ioR 6) }lo vefReer 2012 HG1 ECL: General Emergency Initiating Condition:
HOSTILE ACTION resulting in lo ss of physical control of the facility.
Operating Mode Applicability:
All Examf)le Emergency Action Levels: H 1.1 a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the_ (site specific s.S.ecurity 5h-ift-Shift supervisioASupervisor1. AND b. EITHER of the following has occurred: 1. ANY of the following safety functions cannot be controlled or maintained.
* Reactivity control
* Core cooling [PWR] I R..0 V water level [BWR]
* RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT.
Definitions:
HOSTILE ACTION: An act toward PBNP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air. land. or water using guns, explosives.
PROJECTILEs, vehicles.
or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. terrorism-based EALs should be used to address such activities (i.e .. this may include violent acts between individuals in the owner controlled area). HOSTILE FORCE: One or more individuals who are engaged in a determined assault overtly or by stealth and deception.
equipped with suitable weapons capable of killing, maiming. or causing destruction.
IMMINENT:
The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. PROTECTED AREA: The area under continuous access monitoring and control. and armed protection as described in the site Security Plan. 188 Basis: 189 ~I E I 99 0 I (RevisieR 6) Ne~*eFReer 2012 
}JE;I 99 0 I (Re~*isioA 6) }Joyeffieer 2012 This IC addresses an event in which a HOSTILE FORCE has taken physical contro l of the faci li ty to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions. It a l so addresses a HOSTIL E ACTION leading to a loss of physical control that results in act ual or IMMINENT damage to spent fuel due to 1) damage to a spe nt fuel pool cooling system ( e.g., pumps , heat exchangers , contro l s, etc.) or, 2) l oss of spent fuel pool integrity such that sufficient water level cannot be maintained. Time l y and accurate communications between Security Shift Supervision Supervisor and the Control Room is essential for proper classification of a security-related event. Sec urit y plans a nd terminology are based on the guidance provided by NEI 03-12 , T e mplate for the Security Plan, Training and Qualification Plan , Safeguards Contingency Plan [and Independ ent Spent Fuel Storage Installation Security Program].
Emergency plans and i mplementing procedures are public documents; therefore, EALs shou ld not incorporate Security-sensitive in format i on. This includes information that may be a dvanta geous to a potential adversary , suc h as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Secu ri ty Plan. 190 DeYel0per N0tes: Jl l(;I 99 0 I (Re,*isieA 6) J>l e,*emaer 2012 The (site speeifie seeurity shift supervision) is the title of the on shift individual responsible for supervision of the on shift seeurity foree. Emergeney plans and implementing proeedures are publie doeuments; therefore , EALs should not ineorporate Seeurity sensitive information.
This ineludes information that may be advantageous to a potential adversary , sueh as the partieulars eoneerning a speeifie threat or threat loeation.
Seeurity sensitive information should be eontained in non publie doeuments sueh as the Seeurity Plan. With due eonsideration given to the above developer note , EALs may eontain alpha or numbered referenees to seleeted events deseribed in the Seeurity Plan and assoeiated implementing proeedures. Sueh referenees should not eontain a reeognizable deseription of the event. For eKample , an EAL may be worded as " Seeurity event #2 , #5 or #9 is reported b)' the (site speeifie seeurity shift superYision)." See the related Developer
*Note in Appendi1, B , Definitions , for guidance on the development of a scheme definition for the PROTECTED fu~A. EGL Assignment Attributes:
3 .1.4 .D 191 ECL: General Emergency J>J e l 99 01 (RevisieB
: 6) J>JeYeffieer 20 I 2 HG7 Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.
Operating Mode Applicability:
All 6 EXRmple Emergency Action Levels: 1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. Definitions:
HOSTILE ACTION: An act toward PBNP or its personnel that includes the use of violent force to destroy equipment.
take HOSTAGES.
and/or intimidate the licensee to achieve an end. This includes attack by air, land. or water using guns. explosives.
PROJECTILEs, vehicles.
or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e .. this may include violent acts between individuals in the owner controlled area). IMMINENT:
The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a General Emergency.
192 11 SYSTEM MALFUNCTION ICS/EALS N e l 99 01 (Re&#xa5;i s iOA 6) 1>Jovemaer 2012 Table 8 1: Reeognition Category "8" Initiating Condition Matrix UNUSW .. L EVENT SUl Loss of all offsite AG power eapahility to emergeney huses for 15 minutes or longer. Op. ,\fedcs: 1, 2, 3, ;P-ewer Operetie11, Sf.6l1*htp, Hat S1endh)', Net 8hHtdewn 8U2 illJPLANl>IBD loss of Control Room indieations for 15 minutes or longer. Op. ,\fedes: Pewer Operetien , Stertup , Net 8tffl1dhy*, Net 8J1HtdewnL...1_..
&:1. SID Reaetor eoolant aefrvity greater than Teehnieal Speeifieation allowaele limits. Op. },fades: 1, 2, 3, !fFewer Operetie,"i , 8tertbtf), Net 8t6ll1dhJi, Hat Shutde.~*11 8U4 RCS leakage for 15 minutes or longer. Op. },fades: 1, 2, 3, 4.Pewer Operc1tien, 8te1&deg;tz,.p , Hat Stendhy*, Hat 8/1'1;ltde,n1 SUS Automatie or manual (trip [P\VRJ / seram [BWR]) fails to shutdown the reaetor. Op. },1edes: Pewer Operntie11}
ALERT SAl Loss of all hut one AG power souree to emergeney huses for 15 minutes or longer. Op. Medes: L....2...1,_
1.Pewer Operetien , Sterh1.p, He: 8tclndhy*, Hat 8lmltlew11 8A2 illWLA}l}IBD loss of Control Room indieations for 15 minutes or longer with a signifi e ant transient in progress.
Op. },1edcs: 1, 2, 3. 4 Pewer Operetie11, 8t6lrtup , Hat Stendhy*, Net 8h'l;ltdewn SAS Automatie or manual (trip [PWR] / seram [BWR]) fails to shutdown the reaetor , and suhsequent manual aetions talEen at the reaetor eontrol eonsoles are not sueeessful in shutting down the reaetor. Op. 1',{edes:
Pewer Operetion 1 193 881 SITE AREA EMERGENCY Loss of all off site and al I onsite AG pov ,r er to emer g eney huses for 15 minutes or longer. Op. Jt.1edes:
: 1. 2. 3. 4.Power Operation, 8/ertup , Hat 8t6ll1d/Jy , He: Shutdown sss lnaeility to shutdown the reaetor eausing a ehallenge to (eore eooling [P1&#xa5;R] I R.0 V water leYel [BWJi']) or RCS heat remo\*al. Op. },fades: Pewer 0pa6ltionl 1 GENER4..L EMERGENCY SGl Pro l onged los s of all offsite and all onsite AG power to emergeney huse s. Op. ,\fades: 1. 2. 3. 4.Pewer Operetion , 8t6lrhtp , Hat 8tclndhyi, Hat 8/mtdewn -------------------
: Table intended for use B)' 1 EAL deYelopers.
: lnelusion in lieensee I d . . d , oeuments 1s not require . L------------------'
UNUSUAL EVENT SU6 Loss of all onsite or offsite eommunieations eapabilities.
Op. Mades: 1, 2, 3, 4Pewer Operatien , Steu-htp, Het Stand/:Jj*, Het S/u;1/dew::z SU7 Failure to isolate eontainr-nent or loss of eontainr-nent pressure eontrol. [PWR] Op. Mades: 1, 2, 3, 1. P.ewe1* Operatien , Sta1*titp, Het Standby , Het Shutdewn ALERT SITE AREA. EMERGENCY
}JJ;J 99 0 I (Re,*i s ieR 6) }Je,*effiser 2012 GENERAL EMERGENCY SS8 Loss of all Vital DC SG8 Loss of all AC and SA.9 Hazardous e,.*ent affeeting a SAFETY SYSTEM needed for the eurrent operatiflg mode. Op. J,{edes: 1, 2, 3, 4P-ewer Operatien, Startup , Het Standby*, Het Shutdewn po*,*,cer for 15 minutes or looger. Op. J,{edes: 1, 2, 3, 4Pewer Operatie1:z , Startttp, Het Sta1:zd/:Jj*, He: Sh1;1tdewn 194 Vital DC power sourees for 15 minutes or longer. Op. l',{edes:
1, 2, 3, 4Pewer Operatien, Starthtf), Het Stal:zdhy , Het Sh1;1tdew1:z I-------------------I : Table ifltended for use by 1 EAL developers.
: lnelusion ifl lieensee I d . . d 1 0el:Hflents 1s not require . 1 L------------------J ECL: Notification of Unusual EventUnusual Event ~IEI 99 0 I (RevisieA 6) ~levemeer 2012 SU1 Initiating Condition:
Loss of all offsite AC power capability to emergency buses for 15_-minutes or longer. Operating Mode Applicability:
Po 1 Ner Operation , Startup, Hot Standby , Hot Shutdovm 1. 2. 3. 4 Examf)le Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded , or will likely be exceeded.
S 1.1 Loss of ALL offsite AC power capability to (site specific emergency buses) 1(2)-A-05 and 1(2)-A-06 for 15 minutes or longer. Definitions:
Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses-. .,_This condition represents a potential reduction in the level of safety of the plant. For emergency classification purposes, " capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it. Note: with respect to this EAL. "Sta tion Blackout is Unit 1(2) specific." Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power. Escalation of the emergency classification level would be via IC SAL Develef)er Netes: The " site specific emergency buses" are the buses fed by offsite or emergency AC power sources that supply power to the electrical distribution S)'S tem that powers SAFETY SYSTBMS. There is typically 1 emergenC)' bus per train of SAFETY SYSTEMS. At multi unit stations , the BALs may credit compensatory measures that are proceduralized and can be implemented
\Yithin 15 minutes. Consider capabilities such as power souree cross ties , " sv,'ing" generators , other power sourees described in abnormal or emergency operating procedures, etc. Plants that have a proceduralized capabilit)'
to supply offsite AC po*wer to an 195 
}JE;I 99 0 I (R e vi s i e A 6) 1>Je\*e m'3 e r 201 2 affected unit via a cross tie to a companion unit may credit this pov,cer s ource in the BAL provided that the plaFmed cross tie s trategy meet s the requirements of 10 CFR 50.6 3. EGL Assignment Attribute s: 3 .1.1.A 196 ECL: Notification of Unusual EYentUnusual Event l>IE I 99 0 I (Revi s i e A 6) l>Jev ember 2012 SU2SU3 Initiating Condition:
UNPLANNED lo ss of Control Room indications for 15 minute s or lon ger. Operating Mode Applicability:
Pov,'er Operation , Startup , Hot Standby, Hot Shutdown I. 2, 3 , 4 Example Emergency Action Levels: Note: The Emergency Director s h ou ld declar e the Unusual Eve nt promptl y upon det erm inin g that 15 minutes has been exceeded, or will likely be exceeded. S 3.1 a. An UNPLANNED event results in the inability to monitor one or more of the following parameter s from within the Control Room for 15 minute s or longer.
* Reactor Power
* RCS / Pressurizer Level
* RCS I Pressurizer Pressure
* Core Exit I RCS Temperature
* Level in at least one steam generator
* Steam Generator Auxiliary Feed Water Flow Suppression Pool Temperature Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition.
including the ECCS. These systems are classified as safety-related.
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation i n the level of safety of the plant. 197 l>Jel 99 0 I (Re~*isien 6) l>Je\*ember 2012 As used in this EAL, an " inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).
_For example , the reactor power level cannot be determined from any analog , digital and recorder so urce within the Control Room. 198 l>lB I 99 0 I (Re\*i s i e A *i) l>le veff!eer 2012 An event involving a loss of plant indications, annunciators and/or display sys t ems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.
The event would be reported if it significant l y impaired the capability to perform emergency assessments.
In particular, emergency assessments necessary to implement abnormal operating procedures , emergency operating procedures , and emergency plan implementing procedures addressing emergency classification , accident assessment , or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions ofreactivity control , core cooling [PWR] I RPV level [BWR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.
In addition , if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level [PWR] / RPV water level [BWR] cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer , the availability of other parameter values may be compromised as we! l. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level would be via IC SA iJ_. Developer Notes: IA the PWR parameter list columA , the " site specific Humber" should reflect the miAimum Humber of steam geAerators Aecessary for plaAt cooldowA aAd shutdovm.
This criterioA may also specify whether the level value should be vride raAge , AarrO\'I' raAge or both , depeAdiAg upoA the moAitoriAg requiremeAts iA emergeAcy operatiAg procedures.
Developers may specif y either pressurizer or reactor vessel level iA the PWR parameter columA entry for RCS Le*rel. The Humber , type, locatioA aAd layout of CoAtrol Room iAdicatioAs , aAd the raAge of possible failure modes , caA challeAge the abilit)* of an operator to accuratel)'
determine , withiA the time period available for emergeAcy elassificatioA assessmeAts , ifa specific percentage of iAdicatioAs have beeA lost. The approach used iA this EAL facilitates prompt aAd accurate emergeAcy classificatioA assessmeAts by focusing on the indications for a selected subset of parameters.
By focusing on the availabilit)
* of the specified parameter values , instead of the sources of those values, the EAL recognizes and accommodates the wide variety of indications in nuelear power plant Control Rooms. Indication t)*pes and source s may be analog or digital , safety related or not , primary or alternate , individual meter value or computer group display , etc. A loss of plant anntmciators will be evaluated for reportability in accordance with 10 CFR 50.72 (aAd the associated guidance iA NUREG 1022), and reported if it significantly impairs the capability to perform emergency assessments.
Compensatory measures for a loss of annunciation can be readily implemented and may include increased monitoring of main control boards and more frequent plant rounds by Aon liceAsed operators.
Their alerting functioA notwithstanding , aAnunciators do Rot provide the parameter values or specific compoAent status iAformation used to operate the plant , or process through AOPs or EOPs. Based on these consideratioAs , a loss of 199 
}lei 99 0 I (Re,*isieA 6) "t>loYember 2012 annunciation is considered to be adequately addressed by reportability criteria , and therefore not included in this IC and EAL. With respect to establishing eve0t severity, the response to a loss of radiatioR mo0itori0g data (e.g., process or effluent mo0itor values) is considered to be adequately bounded b)' the requirements of 10 CFR 50.72 (aRd associated guidance iR NUREG 1022). The reporti0g of this eve0t will e0sure adequate plant staff aRd }JRC a*n*areRess , and drive the establishmeRt of appropriate compensatory measures and corrective actioRs. IA addition , a loss of radiation rnoRitoring data , by itself , is not a precursor to a more sigRificant event. Personnel at sites that have a Failure Modes and Effects Analysis (FM.EA) included withiR the design basis of a digital I&C system should consider the FMEA iRformation
,,,*hen developi0g their site specific EALs. Due to cha0ges in the configurations of SAFETY SYSTEMS, i0cludi0g associated instrumentation and indicatioRs, during the cold shutdov,rn , refueling , and defueled modes, RO analogous IC is i0cluded for these modes of operatioR.
EGL AssigRmeRt Attributes:
3 .1.1.A 200 2 l>lE: 1 99 Q I (Revi s i e A 6) l>l e~*emser 2012 SU3SU4 ECL: Notification of Unusual EventUnusual Event Initiating Condition:
Reactor coo lant activity greater than Tec hnical Specification allowable limits. Operating Mode Applicability:
Power Operation, Startup , Hot Standb)', Hot Shutdown 1. 2, 3. 4 Example Emergency Action Levels: ( 1 or 2) (Site specific radiation monitor)Failed Fuel Monitor 1(2)-RE-109 reading greater than (site specific Yalue)750 mR/hr. Sample analy s i s indicate s that a RCS Specific Activity value i s gr e ater than an allowable limit specified in Technical Specifications as indicated by ANY of the following conditions:
: a. Dose Equivalent I-131 greater than 50 &#xb5;Ci/gm b. Dose E quivalent I-1 3 1 greater than 0.5 &#xb5;Ci/gm but le s s than or equal to 50 &#xb5;Ci/gm for great e r than 48 hours c. Dose Equivalent Xe-1 33 greater than 300 &#xb5;Ci/gm for great e r than 48 hour s Sample anal)*sis indicates that a reactor coolant acti1t*ity value is greater than an allov,'able limit specified in Technical Specifications. Definitions:
Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specification s 3.4.16 for longer than the a ll owed comp l etion time. This condition is a precursor to a more significant event and represents a potential degradation of the l evel of safety of the plant. Escalation of the emergency classification level would be via ICs FAI or the Recognition Category A-R ICs. 201 DeYel013er N0tes: N e l 99 O I (Re~*isieA 6) l>Jevemeer 2012 For EAL #1 Enter the radiation monitor(s) that may be used to readily identify 1.vhen RCS activity levels exceed Technical Specification allowable limits. This EAL may be developed using different n~ethods and sites should use eJ&#xa3;isting capabilities to address it (e.g., development of new capabilities is not required).
Examples of eJdsting methods/capabilities include:
* An installed radiation monitor on the letdown system or air ejector.
* A hand held monitor or deployed detector reading with pre calculated conversion values or readily implementable conversion calculation capability. The monitor reading values should correspond to an RCS activity level approximately at Technical Specification allowable limits. If there is no eJdsting method/capability for determining this EAL , then it should not be included.
IC evaluation
\viii be based on EAL #2. For EAL#2 Developers may reword the EAL to include the reactor coolant activity parameter(s) specified in Technical Specifications and the associated allowable limit(s) (e.g., values for dose equivalent I 131 and gross activity, time dependent or transient values , etc.). If this approach is selected , all RCS activity allowable limits should be included. EGL Assignment Attributes:
3 .1.1.A and 3 .1.1.B 202 l>lEI 99 0 I (Revi s i e A 6) l>leve mb e r 2012 SU4SU5 ECL: ~fotification of Unusual EventUnusual E vent Initiating Condition:
RCS leakage for 15 minutes or longer. Operating Mode Applicability:
Po*wer Operation , Startup , Hot Standby , Hot Shutdown 1, 2. 3. 4 Example Emergency Action Levels: (1 or 2 or 3) Note: $2 &3 The Emergency Director should declare the Unusual Event promptl y upon determining that 15 minutes has been exceeded , or will likely be exceeded.
RCS unidentified or pressure boundary leakage greater than (site specific value) 10 gpm for 15 minutes or longer. RCS identified leakage greater than (site specific value)25 gpm for 15 minutes or longer. Leakage from the RCS to a location outside containmen
: t. or St e am G e nerator tube leakage. greater than 25 gpm for 15 minutes or longer. Definitions:
UNISOL A B LE: A n op e n or breached system line that canno t be i s olat e d, remotely or lo c ally. Basis: This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators , following applicable procedures , have been unable to promptly isolate the leak. This condition is considered to be a potential degradation o f the level of safety of the plant. EAL SUS.I and EAL SU5.2 are focused on a loss of mass from the RCS due to " unidentified leakage" , "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications).
EAL SU5.3 addresses a RCS mass loss caused by an UNISOLABLE leak thr o ugh an interfacing system. These EALs thus apply to leakage into the containment , a secondary-side system ( e.g., steam generator tube leakage in a PWR) or a location outside of containment.
The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications.
Lesser values typically require time-consuming calculations t o determine (e.g., a mass balance calculation).
EAL SU5.l uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. 203 NEI 99 01 (Revisien
: 6) l>leYember 2012 The release of mass from the RCS due to the as-designed
/expected operation of a relief valve does not warrant an emergency classification. For PWRsPBNP , an emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected ( e.g., a relief valve sticks open and the line flow cannot be isolated).
For B'NRs, a stuck opeA Safety Relief Valve (SRV) or SRV leakage is Rot coAsidered either ideAtified or uAideAtified leakage by TechAical SpecificatioAs aAd , therefore , is Rot applicable to this EAL. 204 J>JE I 99 0 I (RevisieA 6) J>le vember 2012 The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage , if possible.
Escalation of the emergency classification level would be via ICs of Recognit i on Category A-R or F. DeYelefJeF Netes: EAL #1 For the site specific leak rate valtte , enter the higher of l O gpm or the valtte specified in the site's Technical Specifications for this type of leakage. EAL #2 For the site specific leak rate valtte , enter the higher of 25 gpm or the valtte specified in the site's Technical Specifications for this type of leakage. For sites that have Technical Specifications that do not specify a leakage t)'pe for steam generator rube leakage , developers should inclttde an EAL for tube leakage greater than 25 gpm. for 15 minutes or longer. EGL Assignment Attribtttes:
: 3. l. l .l"r 205 
-------------------------~
----206 }I E;I 99 QI (Re~*isieA 6) }10'1'8A~08F 2Q 12 
}lei 99 0 I (ReYisieA
: 6) 1>leYember 2012 SU5SU6 ECL: Notification of Unusual EYentUnus u a l Event Initiating Condition:
Automatic or manual f trip [P\l/R] / scram [B\l/R]) fails to shutdown the reactor. Operating Mode Applicability:
Power Operationl Nate: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in contro l rods o r imp l ementation of boron injection strateg i es. Example Emergency Action Levels: (1 or 2) Note: A manua l action is any operator act i on. or set of actions, wh i ch causes the control rods to be rapid l y inserted into the core, and does not include manually driving i n control rods or implementation of boron injection strategies.
SU6.l a. An automatic f trip [PWR] I scram [BWR]) did not shutdown the reactor. AND b. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor. a. A manual trip ([P\l/R] / scram [BWR]) did not shutdown the reactor. AND b. EITHER of the following:
Definitions:
: 1. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor. OR 2. A subsequent automatic f trip [P\l/R] / scram [BWRJ) is successful in shutting down the reactor. 207 Basis: }lei 99 0 I (Revision 6) }J o~*emeer 2012 This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor f trip [PWR] I scram [BW~]) that results in a reactor shutdown , and either a subsequent operator manual action taken at the reactor control consoles or an automatic f trip [PWR] I scram [BWR]) is successful in shutting down the reactor. _This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. NOTE: For PBNP, the phrase " at the reactor control consoles" means the reactor trip pushbuttons on the following panels:
* Unit 1 on panels 1C04 and COI
* Unit 2 on panels 2C04 and CO2 208 NB! 99 0 I (Revision
: 6) l>loven1'3er 2012 Following the failure on an automatic reactor f trip [PWR] I scram [BW~]), operators will promptly initi ate manual actions at the reactor control eonsoles in the Control Room to shutdown the reactor (e.g., initiate a manual reactor f trip [PW~] I seram [BWR])). If these manual actions are successfu l in s huttin g down the reactor , core heat generation wi ll quickly fa ll to a l evel within the capabi liti es of the plant's decay heat removal systems. If an initi al manual reactor f trip [PWR] l seram [BWR]) is unsuccessful , operators will promptly take manual act ion at another l ocation(s) on the reactor control conso l es to shutdown the reactor (e.g., initiat e a manual reactor f trip [PWR] I seram [BWR])) using a different switch). Depending upon severa l factors , the initial or subsequent effort to manually f trip [PWR] l scram [BWR]) the reactor , or a concurrent plant condition , may lead to the generation of an automatic reactor f trip [PWR] I seram [BWR]) signa l. If a subsequent manual or automatic f trip [PWR] I seram [BWR]) is successfu l in shutting down the reactor , core heat generation will quickly fall to a l evel within the capabilities of the plant's decay heat removal systems. A manual action at the reaetor eontrol eonsolesin the Control Room i s any operator action , or set of actions , which causes the control rods to be rapidly inserted into the core ( e.g., initiati ng a manual reactor f trip [PWR] I seram [BWR])). This act i on does not include manually driving in control rods or impl ementat ion of boron injection strategies. Actions taken at back-panels or other location s wit hin the Contro l Room , or any loc ation outside the Contro l Room , are not considered to be "at the reactor contro l consoles". Taking the Reaetor Mode Switeh to SHUTDO\\'},J:
is considered to be a manual scram action. {BW/ij The plant response to the failure of an automatic or manual reactor f trip [PWR] / seram [BWR]) will vary based upon severa l factors including the reactor power level prior to the event , availability of the condenser, performance of mitigation equipment and actions , other concurrent plant conditions , etc. If subsequent operator manual actions taken at the reactor control conso l es are also unsuccessful in shutt in g down the reactor, then the emergency classification level will escalate to an Alert via IC SA~. Depending upon the plant response , escalation is also po ssible via IC FAl. Absent the plant conditions needed to meet either IC SA.&sect;.&sect; or FAI , an Unusual Event declaration is appropriate for this event. A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Shou ld a reactor f trip [PWRJ / seram [B'.VR]) signal be generated as a result of plant wo r k ( e.g., RPS setpoint testing), the following classification g uidance should be applied. If the signa l causes a plant transient that should have included an automatic reactor f trip [PWR] I seram [B'.VR]) and the RPS fails to automatica lly s hutdo wn the reactor , then this IC and the EALs are applicable , and should be eva luat ed. I
* If the signa l does not cause a plant transient and the f trip [PWR] / seram [BWR]) failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are n ot applicab l e and no classification is warranted.
209 De>, 1 el0per Netes: J>JI;I 99 0 I (ReYi s ien 6) J>foyerHSer 20 f 2 This IC is applicable in any Mode in which t he actual reactor power level could e~rneed the power level at which the reactor is considered shutdovm. A PWR v,rith a shutdovm reactor power level that is less than or equal to the reactor povrer level which defines the lower bound of Power Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode Applicability. For eM.ample , if the rea c t o r is considered to be s hutdown at 3% and Power Operation starts at >5%, then the IC is al s o applicable in Startup Mode. Developers may include site specific EOP criteria indicative of a successful reactor shutdovm in an EAL statement , the Basis or both (e.g., a reactor power level). The term " reactor control consoles" may be replaced 'Nith the appropriate site specific term (e.g., main control boards). EGL Assignment Attributes:
3 .1.1.A 210 211 ~11;1 99 0 I (RevisieA
: 6) ~J0\'emeer 2012 ECL: Notifica-tioA of UAu s ual EveAtUnusual Event Jl,Je l 99 0 I (R ev i s i e A 6) Ne v e mber 2012 SU6SU7 Initiating Condition:
Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability:
Pov,*er OperatioA , Startup , Hot StaAdb)', H o t ShutdowA 1, 2, 3, 4 Example Emergency Action Levels: (1 or 2 or 3) S 7.1 Lo ss of ALL of th e following on s ite communication methods:
* Plant Public Addres s Sy s tem (Gai-Tronics)
* Commercial Phones
* PBX Phon e s
* Securit y Radio
* Portable Radios S 7.2 Lo ss of ALL o f th e following offsite response organization communications method s:
* Nuclear Accident Reporting System (NARS)
* Commercial Phones
* PBX Phones
* Satellite Phones
* Manito w oc County Sheriffs Department Radio S 7.3 Loss of ALL of th e following NRC communications methods:
* FTS Phon e System
* Commercial Phones
* PBX Phone s
* Satellite Phones Definitions:
None Basis: This IC addresses a significant loss of on-site or offsite communications capabilities.
While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to GR:Qoffsite response organization s and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible ( e.g., use of non-plant , privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points , individuals being sent to offsite locations , etc.). EAL SU7.1 addresses a total loss of the communications methods used in support of routine plant operations.
---EAL SU7.2 addresses a total loss of the communications methods used to notify all GR:Qoffsite response organization s of an emergency declaration.
The GWoffsite response 212 l>J EI 99 0 I (Re~'i s ion 6) November 2012 organization s referred to here are (see Deve l oper Notes)the State of Wisconsin.
Manitowoc County. and Kewaunee County. ---EAL SU 7.3 addresses a total loss of the communications methods used to notify t he NRC of an emergency declaration.
DeYeloper Notes: EAL #1 The " site specifio list of communications methods" should ino l ude all communications methods used for routine plant communications (e.g., commeroia l or site telephones , page party systems , radios , eto.). This listing should inolude installed plant equipment and components , and not items ovmed and maintained by individuals. EAL #2 The " site speoific list of oommunioations methods" should inolude all oommunioations methods used to perform initial emergenoy notifications to OROs as desoribed in the site Emergenoy Plan. The listing should inolude installed plant equipment and oomponents , and not items owAed aAd maiAtained by iAdividuals. Example methods are ring down/dedicated telephoAe lines , oommercial telephone lines , radios , satellite telephones and internet based oommunioation s teohnology
'. In the Basis seotion , insert the site speoifio listiAg of the OROs requiring notificat i on of an emergenoy deolaration from the GoAtrol Room in aocordance
*with the site Emergenoy Plan , and typioally
*within 15 minutes. EAL #3 The "site speoific list of oommunications methods" should include all communications methods used to perform initial emergency notifications to the NRG as described in the site Emergency Plan. The listing should include installed plant equipment and components , and not items owned and maintained by individuals.
These methods are typically the dedicated EmergeAcy Notification System (EN8) telephone line and commercial telephoAe l ines. EGL Assignment Attributes:
3 .1.1.G 213 214 l>Jel 99 O I (ReYisieA
: 6) Ne~*e1tt0er 2012 Nm 99 O I (ReYi s ion 6) ~J ov emser 2012 SU7SU8 ECL: Notification of Unusual Event Initiating Condition:
Failure to isolate containment or loss of containment pressure control. fP-WJij Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdmvn I. 2. 3. 4 Example EmergeeeyEmergency Action Levels: (1 or 2) U8.1 a. Failure of containment to isolate when required by an actuation signal. AND b. ALL required penetrations are not closed within 15 minutes of the actuation signal. a. b. Definitions:
Basis: Containment pressure greater than (site specific pressure)25 psig. AND Less than one full train of (site specific system or equiprnent)Containment Cooling System equipment is operating per design for 15 minutes or longer. This IC addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant. For EAL SU8.l, the containment isolation signal must be generated as the result on an normal/accident condition ( e.g., a safety injection or high containment pressure);
a failure resulting from testing or maintenance does not warrant classification.
The determination of containment and penetration status -isolated or not isolated -should be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations , if possible.
EAL SU8.2 addresses a condition where containment pressure is greater than t he setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. During a design basis accident.
a minimum of two containment accident fan cooler units with their accident fans running and one containment spray train are required to maintain the containment peak pressure and temperature below the design limits. Each containment spray train is a containment spray pump, spray header. nozzles. valves and piping. Each containment accident fan cooler unit consists of cooling coils. 215 
?>161 99 0 I (ReYisieR
: 6) ?>Jeyeffiaer 2012 accident backdraft damper. accident fan, service water outlet valves, and controls necessary to ensure an operable service water flow path. 216 217 }lei 99 0 I (Re\*isiefl 6) }l0\'eff\eer 20 J 1 J>Je l 99 0 I (Re\*i s ieA 6) }le\*e11'leer 2012 The 15-minute criterion is inc l uded to allow operators time to manually start equipment that may not have automatically started, if possible.
The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fan.st-are either lost or p erfo rming in a degraded manner. This event would escalate to a Site Area Emergency in accordance with IC FS 1 ifthere were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.
De-;el0per N0tes: Enter the " site specific pressure" value that actuates containment pressure control systems (e.g., containment spray). Al s o enter the site specific containment pressure control system)equipment that should be operating per design if the containn,ent pre ss ure actuation setpoint is reached. If desired , specific condition indications such a s parameter value s can al s o be entered (e.g., a containment spray flov ,r rate less than a certain value). EAL #2 is not applicable to the U.S. Evolutionary Power Reactor (EPR) design. EGL Assignment Attributes:
3 .1.1.A 218 219 l>H':l 99 0 I (ReYisieA
: 6) l>le*,emeer 2012 I ~------J>J E;I 99 0 I (Revi s ieA 6) Nevember 2012 SA1 ECL: Alert Initiating Condition:
Loss of all but one AC power source to emergency buses for 15 minutes or longer. Operating Mode Applicability:
Pov,'er OperatioA , Startup , Hot StaAdby , Hot Shutdovm I. 2. 3. 4 Example EmergeneyEmergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15_ minutes has been exceeded, or will likely be exceeded.
fil l a. AC power capabi li ty to (site specific emergeAcy buses) I (2)-A-05 AND I (2)-A-06 is reduced to a single power source for 15 minutes or longer. AND b. Any additional single power source failure will result in a loss of a+l-ALL AC power to SAFETY SYSTEMS. Definitions:
SAFETY SYSTEM: A system required for safe plant operation.
cooling down the plant and/or placing it in the cold shutdown condition.
including the E CCS. These sy s tems are classified as safety-related. Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition , the sole AC power source may be powering one, or more than one , train of safetyrelated equipment.
This IC provides an esca lati on path from IC SUL
* Normal Unit 1(2) offsite power sources include: o 345 KVAC 1(2)X-03 through the 13.8 KVAC system to the LVSAT 1(2)X-04 o 3 45 KV AC backfed through the 19 KVAC sy s tem to the UAT 1 (2)X-02
* Normal Unit 1(2) onsite power sources consist of: o emergency diesel generators o gas turbine generator o unit main turbine generator o power supplied from the opposite unit 220 l'JB I 99 0 I (Revi s i o R 6) l'lov effis e r 2012 An "AC power source" is a source recognized in AOPs and EOPs (including Beyond De s ign Basi s event procedures), , and capable of supplying required power to an emergency bus. Some examples of this Initi a ting Ce ondition are presented below.
* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
* A loss of emergenc y power sources ( e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source. 221 222 l>lel 99 O I (Re~*isieR
: 6) l>le~*emeer 2012 
~l e i 99 O I (Re,*isieA
: 6) ~Je,*effieer 2012 Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SSL Develeper Netes: For a pov,er source that has multiple generators , the EAL and/or Basis section shou l d reflect the minim.um number of operating generators necessary for that source to provide required po1,yer to an AC emergenC)'
bus. For eJrnmple, if a backup po*Ner source is comprised of two gene r ators (i.e., hvo 50% capacity generators sized to feed 1 AC emergency bus), the EAL and Basis section must specify that both generators for that source are operating. The "site specific emergency buses" are the buses fed by offsite or emergenC)'
AC power sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typically 1 emergency bus per train of SAFETY SYSTEMS. Developers should modify the bulleted eJrnmples prov i ded in the basis section, above, as needed to reflect their site specific plant designs and capabilities.
The EALs and Basis should reflect that each independent offsite power circuit constitutes a single power source. For eJrnmple, three independent 345kV offsite power circuits (i.e., incoming power lines) comprise three separate power sources. Independence may be determined from a review of the site specific UFSAR , SBO analysis or related loss of electrical power studies. The E,c\L and/or Basis section may specify use of a non safety related power source provided that operation of this source is recognized in AOPs and EOPs, or beyond design basis accident response guidelines (e.g., FLEX support guidelines).
Such power sources shou l d generally meet the " Alternate ac source" definition proYided in 10 CFR 50.2. At multi unit stations , the EALs may credit compensatOf)
' measures that are proceduralized and can be implemented vt'ithin 15 minutes. Consider capabilities such as power source cross ties, "s wing" generators, other power sources described in abnormal or emergency operating procedures , etc. Plants that haYe a proceduralized capability to supply offsite AC power to an affected unit Yia a cross tie to a companion unit may credit this power source in the EAL provided that the planned cross tie strategy meets the requirements of l O CFR 50.63. EGL Assignment Attributes:
3 .1.2.B 223 ECL: Alert }JB I 99 0 I (Revi s i e H 6) N o vember 2012 SA2SA3 Initiating Condition:
UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
Operating Mode Applicability:
P01,ver Operation , Startup , Hot Standby , Hot Shutdown I. 2, 3. 4 Example Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15 _ minutes has been exceeded , or will likely be exceeded.
S 3.1 a. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.
* Reactor Power
* RCS / Pressurizer Level
* RCS / Pressuri z er Pre ss ure
* Core Exit/ RCS Temperature
* Levels in at least one st e am generator
* Steam Generator A u x iliary F eed Water Flow Suppression Pool Temperature AND =b*;..__----'
ANY of the following transient events in progress.
* Automatic or manual run back greater than 25% thermal reactor power
* Electrical load rejection greater than 25% full electrical load
* Reactor scram [BWR] I trip [PWR]
* EGGS (SI} actuation
* Thermal pov,er oscillations greater than 10% [BWR] Definitions:
SAFETY SYSTEM: A system required for safe plant operation.
cooling down the plant and/or placing it in the cold shutdown condition.
including the ECCS. These systems are c l assified as safety-related.
224 
}l e i 99 0 I (RevisisA e) }le\*ember 20 I? UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant condi t ions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition , the margin to a potential fiss i on product barrier challenge is reduced. It thus represents a potential s ub stantial degradation in the l evel of safety of the plant. 225 1>Je l 99 0 I (R e\*i s i o n 6) 1>Jove A'll3 e r 2012 As used in this EAL , an " inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all o f the Control Room sources for the given parameter(s).
For example , the reactor power level cannot be determined from any analog , digital and recorder source within the Control Room. An event involving a loss of plant indications , annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.
The event would be reported if it significantly impaired the capability to perform emergency assessments.
In particular , emergency assessments necessary to implement abnormal operating procedures , emergency operating procedures , and emergency plan implementing procedures addressing emergency classification , accident assessment , or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling [PWR] I R_.DV le*rel [BWR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition , if all indication sources for one or more of the listed parameters are lost , then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example , if the value for reactor vessel level [PWR] / RP\' 'Nater level [BWR] cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level would be via ICs FSI or IC M+RS 1. De7,releper Netes: IA the P'.l/R parameter list columA , the " site s pecific Aumeer" should reflect the miAimum Aumeer of steam generators necessary for plant cooldown and shutdown.
This criterion may also specify whether the level value should ee 1 Nide raAge , narrow range or eoth , depending upon the monitoring requirements in emergenC)' operating procedures.
Developers may specify either pressurizer or rea c tor vessel level in the PWR parameter column entry for RCS Level. Developers should consider if the " transient event s" list needs to ee modified to setter reflect site specific plant operating characteristics and expected respoAses.
The numeer , type , location and layout of Control Room iAdicatioAs , and the range of possible failure modes, can challenge the ability of an operator to accurately determiAe , vl'ithin the time period available for emergency classification assessments , if a specific percentage of indications have been lost. The approach used in this EAL facilitates prompt and accurate emergency classification assessmeAts by focusing on the iAdications for a selected subset of parameters.
By focusing on the availability of the specified parameter values , instead of the sources of those values, the EAL recognizes and accommodates the wide variety of indications in nuclear power plaAt Control Rooms. Indication types and sources may be analog or digital , safety related or not , primary or alternate , individual meter value or computer group display , etc. 226 l'J e l 9 9 0 I (Revi s i o A 6) }JoveA'!ser 2012 i\ loss of plant annunciators
*.viii be evaluated for reportability in accordance with 10 CFR 50.72 (and the associated guidance in NUREG 1022), and reported if it s ignificantly impairs the capability to perform emergency as s essments. Compensatory mea s ures for a loss of a11Hunciation can be readily implemented and may include increased monitoring of main control boards and more frequent plant rounds by non lic e n s ed operators. Their al e rting function nohNithstanding , annunciators do not provide the parameter values or specific comp o nent status information used to operate the plant, or process through AOPs or EOP s. Based on the s e considerations , a l o ss of annunciation is considered to be adequately addressed by reportability criteria , and therefore not included in this IC and EAL. \Vith respect to establishing event s everity , the response to a los s of radiation monitoring data (e.g., proces s or effluent monitor value s) is considered to be adequately bounded by the requirements of 10 CFR 50.72 (and associated guidan c e in NUREG 1022). The reporting of this event will ensure adequate plant staff and NRG 8:\Varene ss, and drive the establi s hment of appropriate compensatory measure s and corrective actions. In addition , a loss of radiation monitoring data , by itself , is not a precursor to a more significant event. Personnel at sites that have a Failure Modes and E f fe c ts Anal y sis (H,4EA) included *.vithin the design basis of a digital l&C system should consider the FMEA information when developing their site specific EALs. Due t o changes in the configurations of SAFETY SYST E MS , including associated instrumentation and indications , during the cold shutdown , refueling , and defueled modes , no analogous IC is included for these modes of operation. EGL Assignment Attributes
: 3 .1.2.B 22 7 228 M;;I 99 0 I (RevisieR
: 6) ~leYeR10er 2012 ECL: Alert J>JE;I 99 0 I (RevisieA
: 6) l>l0YeA10er 2012 SA5SA6 Initiating Condition:
Automatic or manual f trip [PWR] / scram [BWR]) fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successfu l in shutting down the reactor. Operating Mode Applicability:
Power Operationl Note: A manual action is any operator action, or set of actions, which causes the contro l rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
J Examf)le EmergeneyEmergency Action Levels: .s_k 1 a. An automatic or manual f trip [PWR] / scram [BWR]) did not shutdown the reactor. AND b. Manual actions taken at the reactor control consoles are not successful in sh utting down the reactor. Definitions:
Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor f trip [PWR] I scram [BWR]) that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also un successful.
This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS. NOTE: For PBNP, the phrase "at the reactor control consoles" means the reactor trip pushbuttons on the following panels:
* Unit 1 on panels 1 C04 and CO 1
* Unit 2 on panels 2C04 and CO2 A manual action at the reactor control consoles is a ny operator action, or set of actions, wh ich causes the control rods to be rapidl y inserted into the core (e.g., initiating a manual reactor f trip [PWR] I scram [BWR])). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsucce ssful, operators wo uld immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers).
_Actions taken at back-panels or other locatio ns within t h e Control Room , or any location outside the Control Room, are not considered to be " at the reactor control consoles~''-=-229 
:t>le l 99 0 I (R e vi s i o A 6) N o v e A~ber 2012 Taking the Reactor Mode Switch to SHUTDO'.l/tl is considered to be a manual scram action. [BWR] The plant response to the fai lur e of an automatic or manual reactor f trip [PWR] I scram [BWR]) will vary based upon several factors including the reactor power level prior to the event , avai l ability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions , etc. If the fa ilur e to shutdown the reactor is prolonged enough to cause a challenge to the core cooling [PWR] / R12V v,ater level [B\l/R] or RCS heat removal safety functions , the emergency classification l eve l will escalate to a Site Area Emergency via IC SS~&sect;_. Depending upon plant responses and symptoms , escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either IC SS~&sect;_ or FS 1 , an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however , this IC and EAL are included to ensure a timely emergency declaration.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
De>;elef)er Notes: This IC is applicable in any Mode in which the actual reactor pov,rer level could exceed the power leYel at which the reactor is considered shutdown.
A PWR with a shutdO\lt'A reactor power leYel that is less than or equal to the reactor pov,'er level which defines the lower bound of Power Operation (Mode 1) will Heed to iAclude Startup (Mode 2) in the OperatiAg Mode Applicability.
For e~rnmple , if the reactor is considered to be shutdowA at 3% aAd Pov,er OperatioA starts at >5%, theA the IC is also applicable in Startup Mode. Developers may iAclude site specific EOP criteria indicative of a successful reactor shutdowA iA aA EAL statemeAt , the Basis or both (e.g., a reactor power level). The term. " reactor coAtrol coAsoles" may be replaced with the appropriate site specific term (e.g., maiA coAtrol boards). EGL AssignmeAt Attributes:
3.1.2.B 230 231 }lei 99 01 (Re\*isieA
: 6) }l0\'eA1ber 20 I?
N E I 99 0 I (RevisieR
: 6) 1>Je~*effiser 2012 SA9 ECL: Alert Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability:
Pov,'er Operation , Startup , Hot Standby , Hot Shutdovm I. 2. 3. 4 Example Emergency Action Levels: Notes: fil l
* If the affected SAF E TY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
* If the hazardous event only resulted in VISIBLE DAMAGE. with no indications of degraded performance to at least one train of a SAFETY SYSTEM. then this emergency classification is not warranted.
: a. b. The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
.!__(site specific hazards)Lake level greater than or equal to +9.0 ft. (Plant elevation)
* Pump bay level less than -19 .0 ft.
* Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director AND 1. 2. E vent damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode. AND EITHER of the fo ll owing:
* Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode. or
* The event has resu l ted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. 232 Definitions:
}J E;I 99 0 I (RevisieR
: 6) }Jevember 2012 EXPLOSION:
A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding.
arcing. etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping dr i ve belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
SAFETY SYSTEM: A system required for safe plant operation.
cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Damage resulting from an equipment failure and limited to the failed component (i.e .. the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. EITHER of the followittg:
: 1. BYeAt damage has caused iAdieatioAs of degraded performaAce iA at least oAe traiA ofa SAFETY SYSTEM Reeded for the curreAt 013eratiAg mode. OR 2. The evettt has caused VISIBLE DAM, 1\GE to a SAFETY SYSTEM compottettt or structure Heeded for the currettt 013eratittg mode. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train. and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words. in order for this EAL to be classified, the hazardous event must occur. at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria SA9. l.b.1 of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded 233 T>Jel 99 O I (Re~*isieA 6) T>levemeer 2012 performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. 234 
}le i 99 0 I (R e vi sie A 6) }leYe A10 e r 2012 VISIBLE DAMAG E addre s ses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will m a ke this determination based on the totality of avai l able event and damage report information.
This is intended to be a brief a s se s sment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAG E should be significant enough to cause concern regarding the operability or reliability of the SAF E TY SYSTEM train. 235 Nel 99 01 (Revisien
: 6) J>JeyerRBeF 2012 This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the High lake water level conditions that may have resulted in a plant VITAL AREA being subjected to levels beyond design limits. and thus damage may be assumed to have occurred to plant SAFETY SYSTEMS. Lake water level at +9.0 feet corresponds to the license basis flood elevation and is one foot above the Turbine Building floor elevation.
The low pump bay level setpointthreshold corresponds to the leve l that is calculated to correspond to the onset of cavitation of the service water pumps. EAL l.b.l addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. EAL l .b.2 addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.
Operators
*v1ill make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC FS 1 or A-8-lRS 1. De*;elaper Nates: For (site specific hazards), de*,celopers should consider including other significant, site specific hazards to the bulleted list contained in EAL I .a (e.g., a seiche). Nuclear power plant SAFETY SYSTEMS are comprised of two or more separate and redundant trains of equipment in accordance with site specific design criteria.
EGL Assignment Attributes:
_3.1.2.B 236 l'JE I 99 0 I (Re~*i s i e A 6) Neve mber 2012 551 ECL: Site Area Emergency Initiating Condition:
Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown 1, 2, 3. 4 Example Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
S 1.1 Loss of ALL offsite and ALL onsite AC power to (site specific emergency buse s) A-05 and 1(2)-A-06 for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. Basis: For the purpose of cl as sification under this EAL evaluation of power sources should b e made on each unit individually. This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control , spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Consideration should be given to operable loads necessary to remove decay heat or prov i de Reactor Vessel makeup capability when evaluating loss of AC power to safety-related 4 1 60 V AC busses. Even though a safety-related 4160 VAC bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not operable on the energized bus then the bus should not be considered operable.
If mitigative strategie s establish emergency power to any bus listed in the E AL the EAL threshold for this Initiating Condition is not met. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via ICs AG+RG 1 , FG 1 or SG l. 237 Develeper Netes: 1>JE;1 99 0 I (RevisieA 6) J>Je yemser 20 J 2 For a po1,ver source that has multiple generators , the EAL and/or Basis section shou l d reflect the minimum number of operating generators necessary for that source to provide adequate power to an AC emergency bus. For e~(ample , if a backup po*.ver source is comprised of two generators (i.e., t\vo 50% capacity generators sized to feed 1 AC emergency bus), the EAL and Basis sect i on must specify that both generators for that s ource are operating.
The " site specific emergenCJ' buses" are the bu s es fed by offsite or emergenCJ' /\C power sources that supply power to the electrical distribution sy s tem that powers SAFETY SYSTEMS. There i s typically 1 emergency bus per train of SAFETY SYSTEMS. The EAL and/or Basis section may speci fy u s e of a non safety related power source provided that operation of thi s source is controlled in ac c ordance with abnormal or emergency operating procedure s, or beyond design basis accident response guidelines (e.g., FLEX support guidelines). Such pmver sources should generally meet the " Alternate ac source" definition provided in 10 CFR 50.2. At multi unit stations , the EALs may credit compensatory measures that are proceduralized and can be implemented within 15 minutes. Consider capabilities such as pov,'er source cross ties , " swing" generators, other power sources described in abnormal or emergency operating procedures , etc. Plants that have a proceduralized capability to supply offsite AC power to an affected unit Yia a cross tie to a companion unit may credit thi s power source in the EAL provided that the planned cross tie strategy meets the requirements of 10 CFR 50.63. EGL Assignment Attributes:
3.1.3.B 238 
}IB I 99 0 I (Re~*isieA 6) }leY eA10er 2012 ECL: Site Area Emergency Init i ating Condition: Loss of all Vita l DC power for 15 minutes or longer. Operating Mode Applicability:
: 1. 2. 3. 4 Emergency Action Levels: Note: The Emergency Director should dec l are the Site Area Emergency promptly upon determining that 15 minutes has been exceeded.
or will likely be exceeded. SS2 S 2.1 Indicated volta e i s less than 115 VDC on ALL Vital DC bus ses 0-01. D-02. D-03. and D-04 for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A system required for safe plant operation. cooling d o wn the plant and/or plac i ng it in the cold s hutdown condition, including the ECCS. These s y s tems are cla s sified as safety-related. This IC addresse s a lo ss of Vital DC power which compromise s th e ability to monitor and control SAFETY SYSTEMS. In modes above Co l d Shutdown.
thi s condition involves a ma j or failure of p l ant functions needed for the protection of the public. Fifteen minutes was selected as a thresho l d to exclude transient or momentary power losses. Escalation of the emergenc y classification level wou l d be via I C s RG 1, FG 1 or SG2. 239 
}le i 99 0 I (Revi s ieA 6) }leve mb e r 2012 SS5SS6 ECL: Site Area Emergency Initiating Condition:
Inability to shutdown the reactor causing a challenge to f core cooling [PWR] I R.0 V 1.vater level [BWR]) or RCS heat removal. Operating Mode Applicability:
Po\ver Operationl Example Emergency Action Levels: S 6.1 a. An automatic or manual f trip [PWR] / scram [BWR]) did not shutdown the reactor. AND b. All manual actions to shutdown the reactor have been unsuccessful.
AND c. EITHER of the following conditions exist: * (Site specific indication of an inability to adequate!)
' remoYe heat from the oorejConditions requiring entr y into Core Cooling -Red Path {CSP-C.1) are met. * (Site specific indication of an inability to adequately remove heat from the RGS)Conditions requiring entry into Heat Sink -Red Path {CSP-H.1) are met. Definitions:
Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor f trip [PWR] I scram [BWR]) that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful , and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.
In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor. A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
240 
}16 1 99 QI (RevisieR 6) }l e~*effieer 2012 Escalation of the emergency classification level would be via IC AG+-RG I or FG 1. De'lel0f)er N0tes: This IC is applicable in any Mode in which the actual reactor povt'er level could e1weed the pov,cer le*,cel at which the reactor is considered shutdown.
1\ PWR 1.&#xa5;ith a shutdown reactor power level that is less than or equal to the reactor power le*,*el *n*hich defines the lo*..,.er bound of Po*,ver Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode Applicability.
For e1,ample , if the reactor is considered to be shutdown at 3% and Power Operation starts at >5%, then the IC is also applicable in Startup Mode. Developers may include site specific EOP criteria indicative of a successful reactor shutdown in an EAL statement, the Basis or both (e.g., a reactor power level). Site specific indication of an inability to adequately remove heat from the core: [BWR] Reactor vessel *n<ater level cannot be restored and maintained above Minimum Steam Cooling R.0 V Water Level (as described in the EOP bases). [PWR] lAsert site specific values for an incore/core e1dt thermocouple temperature and/or reactor vessel water level that drives entry into a core cooling restoration procedure (or othenvise requires implementation of prompt restoration actions).
Alternately, a site may use incore/core exit thermocouple temperatures greater than 1 , 200&deg;P and/or a reactor vessel , Y ater level that corresponds to apprmdmately the middle of acfr,e fuel. Plants with reactor vessel level instrumentation that caftllot measure down to apprmdmately the middle of acfr,ce fuel should use the lovi<est on scale reading that is not abo*,e the top of active fuel. If the lowest on scale reading is above the top of active fuel , then a reactor vessel level value should not be included.
For plants that have implemented Westinghouse Owners Group Emergency Response Guidelines , enter the parameters used in the Core Cooling Red Path. Site specific indication of an inability to adequately remove heat from the RCS: [BWR] Use the Heat Capacity Temperature Limit. This addresses the inability to remove heat via the main condenser and the suppression pool due to high pool water temperature.
[PWR] Insert site specific parameters associated with inadequate RCS heat removal via the steam generators.
These parameters should be identical to those used for the Inadequate Heat Removal threshold Fuel Clad Barrier Potential Loss 2.B and threshold RCS Barrier Potential Loss 2.A in the PWR EAL Fission Product Barrier Table. EGL Assignment Attributes
: 3 .1.3 .B 241 8&#xa3;8 ECL: Site Area Emergency Ieitiatieg CeeditieB:
Loss of all Vital DC pov,*er for 15 min1:1tes or longer. T>JE;J 99 0 I (Revi s ien 6) T>levemb e r 2012 Operatieg Mede A.pplieahility:
Power OperatioA , Start1:1p , Hot Standby , Hot ShutdovmL 2, 3, 4 Example Emergeeey Aetiee LeYels: Nete: The Emergeacy Director should declare the Site Area Emergeacy promptly upon determiaing that 15 minutes has beea exceeded , or *will likely be eJweeded.
1 ladicated voltage is less thaa (site specific bus voltage value) 115 VDC 00 A.LL (site specific Vital DC busses)1(2)
D 01, D 02, D 03, aad D 04 for 15 miautes or loager. Basist SAFETY SYSTEM: A system required for safe plaat operatioa , cooling dovm the plaat aad/or placing it ia the cold shutdowa coadition, iacludiag the EGGS. These Sj'Stems are classified as safety related. This IC addresses a loss of Vital DC po*wer whi c h c ompromises the abilitj* to moaitor aad coatrol SA..1''ETY SYSTEMS. la modes aboYe Cold Shutdowa , this coaditioA iavolves a major failure of plaat functions Reeded for the protectioa of the public. Fifieea miautes 1.Yas selected as a threshold to eM.c lude traasieat or momeatary power los s e s. Escalation of the emergency classificatioa level *N ould be via ICs AGlRQl , FGl or SG8. DeYeleper Netes: The "site specific b1:1s Yoltage value" should be ba s ed on the minimum bus voltage necessary for adequate operatioH of SAFETY SYSTEM equipmeat.
This voltage value sho1:1ld iHcorporate a margiH of at least 15 miautes of operatioA befor e the on s et of inability to operate those loads. Thi s voltage is usually Rear the miaimum Yoltage selected wheA battery siziag is performed.
The tj*pical value for an eatire battery set is apprm(imately 105 VDC. For a 60 cell striag of batteries , the cell Yoltage is apprm1.imately 1.75 Volt s per cell. For a 58 striag battery set , the minimum voltage is apprmdmately 1.81 Volts per cell. The " site specific Vital DC busses" are the DC bu sses that provide monitoring and control capabilities for SAFETY SYSTEMS. E GL AssigHment Attributes
: 3 .1.3 .B 242 
}JBI 99 0 I (Revision
: 6) }J0>,en1ber 2012 SG1 ECL: General Emergency Initiating Condition:
Prolonged los s of all offsite and all onsite AC power to emergency buses. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdovm 1. 2. 3. 4 ExamJJle Emergency Action Levels: Note: ~I The Emergency Director should declare the General Emergency promptly upon determining that (site specific hours)4 hours has been exceeded , or will likely be exceeded.
: a. Loss of ALL offsite and ALL onsite AC power to (site specific emergency 005e5fl (2)-A-05 and 1(2)-A-06. AND b. EITHER of the following:
* _Restoration of at least one AC emergency bus in less than (site specific hours)4 hours is not likely.
* _(Site specific i0dication of an inability to adequately remove heat from the eeretCondit i ons requiring entry into Core Cooling -Red Path (CSP-C.1) are met. Definitions:
SAFETY SYSTEM: A system required for safe plant operation.
cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: For the purpose of classification under this EAL evaluation of power sources should be made on each unit individually.
This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core coo lin g , containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged l oss of these buses will lead to a loss of one or more fission product barriers.
In addition , fission product barrier monitoring capabilities may be degraded under these conditions. If mitigative strategies establish emergency power to any bus I isted in the EAL the EAL threshold for this Initiating Condition is not met. The EAL sho uld require declaration of a General Emergency prior to meeting the thresholds for IC FG 1. This wi ll allow addit ion a l time for implementation of offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station 243 1>JE I 99 0 I (RevisieA
: 6) 1>levem13er 2012 blackout coping period. Beyond this time , plant re s ponses and event trajectory are subject to greater uncertainty , and there is an increased likelihood of challenges to multiple fission product barriers.
244 l>Jel 99 01 (ReYisieA
: 6) }Je;*emeer 2012 ---The estimate for restoring at least one emergency bus should be based on a realistic appraisal of the situation.
Mitigation actions with a low probability of success shou ld not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for , and implement , protective actions for the public. 245 
~l e i 99 0 I (Revi s ieA 6) Nevemser 2012 The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core. DeYelepeF Netes: Although this IC and EAL may be viewed as redundant to the Fission Produet Barrier ICs , it is ineluded to provide fur a more timely esealation of the emergeney elassifieation level. The "site speeifie emergeney buses" are the buses fed by offsite or emergeney AC po*.ver sources that supply power to the eleetrical distribution system that pov,'ers SAFETY SYSTEMS. There is typieally 1 emergene;' bus per train of SAFETY SYSTEMS. The "site speeifie hours" to restore AC power to an emergeney bus should be based on the station blackout coping analysis perfurmed in aeeordance with IO CFR &sect; 50.63 and Regulatory Guide 1.155, Slcttien Bl-aelwu:.
Site speeifie indieation ofan inability to adequately remove heat from the eore: [BWR] Reaetor vessel water level eannot be restored and maintained above Minimum Steam Cooling R.0 Y 'Nater Level (as described in the EOP bases). [PWR] lAsert site speeifie 1 ,calues for an ineore/eore e>Lit thermocouple temperature and/or reaetor vessel 'Nater level that drive entry into a eore cooling restoration proeedure (or otherwise requires implementation of prompt restoration aetions).
Alternate);, , a site may use ineore/eore exit thennoeouple temperatures greater than 1 , 2006F and/or a reaetor vessel 1 Nater level that eorresponds to apprmdmately the middle of aetive fuel. Plants 1 ,vith reaetor vessel level instrumentation that eannot measure down to apprmdmately the middle of active fuel should use the lowest on seale reading that is not above the top of active fuel. If the lovt'est on seale reading is above the top of actiYe fuel , then a reaetor yessel level value should not be ineluded.
For plants that hai,e implemented Westinghouse Owners Group Emergeney Response Guidelines , enter the parameters used in the Core Cooling Red Path. EGL Assignment Attributes:
3 .1 A .B 246 
~I E I 99 0 I (Revi s i a A 6) ~1 9\'emeer 2012 SG8SG2 ECL: General Emergency Initiating Condition:
Loss of all AC and Vital DC power sources for 15 minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown I. 2. 3. 4 Examf}le Emergency Act ion Levels: Note: The Emergency Director should declare the Genera l Emergency promptly upon determining that 15 minutes has been exceeded , or wil l likel y be exceeded.
~l a. b. Definitions:
Loss of ALL offsite and ALL onsite AC power to (site specific emergency
---bRll+-ts~es
..... )1(2}:A-05 and 1(2)-A-06 for 15_-minutes or longer. AND Indicated voltage is less than (site specific bus Yoltage value) 115 VDC on ALL (site specific Vital DC busses) D-0 I, D-02. D-03 and D-04 for 15 minutes or lon ger. SAFETY SYSTEM: A system required for safe plant operation.
cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of a ll AC power compromises the performance of all SAFETY SYSTEMS requiring electric power includin g those nec essary for emergency core cooling , containment h eat removal/pressure control , spent fuel heat removal and the ultimat e heat s ink. A lo ss of Vital DC power compromises the ability to monitor and contro l SAFETY SYSTEMS. A s u stained lo ss of both AC and DC power w ill lead to multiple challenges to fission product barriers.
If mitigative strategies establish emergency power to any bus listed in the EAL. the EAL threshold for this Initiating Condition is not met. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-m inu te emergency declaration clock begins at the point when both EAL thresholds are met. The " site specific emergency buses" are the buses fed by offsite or emergency AC po*wer sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typically 1 emergency bus per train of SAFETY SYSTEMS. The " site specific bus voltage Yalue" should be based on the minimum bus voltage necessary for adequate operation of SAFETY SYSTEM equipment.
This *,oltage value should 247 
~m l 99 0 I (RevisieA 6) ~l Member 2012 incorporate a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is usually near the minimum voltage selected when battery sizing is performed.
The typical value for an entire battery set is apprmdmately 105 VDC. For a 60 cell string of batteries , the cell voltage i s approximately 1.75 Volts per cell. For a 58 string battery set , the minimum voltage is approximately 1.81 Volt s per cell. The " site specific Vital DC bu ss es" are the DC busses that provide monitorin g and control capabilities for SAFETY SYSTEMS. This IC and EAL were added t o Revi s ion 6 to addre s s operating eJ , perience from the March , 2011 accident at Fukushima Daiichi. EGL Assignment Attributes:
3.1.4.B 2 48 l>Je l 99 01 (R e Yi s i e A 6) }leYe FAeer 2012 APPENDIX A -ACRONYMS AND ABBREVIATIONS AC ......................................................................................................................
Alternatin g Current AOP ............
...............................
.....................
.................................
Abnormal Operating P r ocedure A ...*******..*.....................................................*.*..*.*..*...........*..*.......*****..*****.***.....*.**.*********..** I'i PR,\4 ..............
...................................................................................... Average Po v , , er Range ~4eter A TWS ................................................................................... Anticipated Transient Without Scram .......................................................................................................................................... B &'.V ................................................................................................................... Babeoek and '.V il eox .......................................................................................................................................... B IIT ....................................................................................... Boron lnjee t ion Initiat i on Temperature
.......................................................................................................................................... B '.V R ...................
............................................................................................. Bo ilin g '.\later Reaetor CDE .........................................................................
......................
....... Committed Dose Equivalent CFR ..............................
........................................
.................
............... Code of Federal Regulations CTh1T/CNMT .............................
................................
...........................
....................... Containment CSF ....................................................................
......................................... Critical Safety F unction ...................................................................................................
....................................... C SFST ................................................
........................................ Critica l Safety Function Status Tree D Bi\ .............................................................................................................. Des i g n Basis 1\eeident DC ................
............................
......................
............................................................ Direc t Current EAL ...........................................
.....................
...................
........................ E mergency Action Level ECCS ................
............................................................................ E mergency Core Coolin g System ECL ..............................
.................................................
.................
Emergency C la ssification Level EOF ...........
............
........................................................................... Emergency Operation s Facility EOP ........................
.............................................
.......................... Emergency Operating Procedure EPA ...........................................................
............
...................... Env ironm ental Protection Agency EPG ..........................
...............
................
...................................... Emergency Procedure Guideline
.......................................................................................................................................... E PIP ................................................................................... Em.ergeney P l an Imp l e m enting Proeedure
.......................................................................................................................................... E P R ......................................................................................................... Evo lu tio nar y Po 1.ver Re a etor .......................................................................................................................................... E PRI .............................................
.............
.........................
............ E l eetr i e Powe r R eseareh Inst i tute .................................................................
......... : ...........................................
.................... E RG ...............................................................
................................... Eme r geney R es p o n se Gu id e li ne .......................................................................................................................................... F EMA ................................
.....................
...............
............ Federal E mergency Management Agency F8AR ..........
....................................................................................
..... F i na l 8 a fe t)' A n a l ysis Report GE .........................................
..................
........................................................... General E mergency ...................................
............
........................................................................................... H CTL ..................
........................................................................... Heat Capae i ty Te mp eratu r e Lim i t .......................................................................................................................................... H P CI ...............................................................................
..................
H i g h Press u re Coo l ant In j eet i on .......................................................................................................................................... H 8I ............
.....................................
...................................................
............ Huma n System Interfaee N e l 99 O I (R e visieH 6) }Jevemeer 2012 IC .....................................
............................................................................
....... Initiating Condit ion ......................................................................................................
.................................... I D ......................................................................................................
......................... In s id e Di a m eter IPEEE. ............................ Individual Plant E>(amiaation of E>{temal Events (Generio Letter 88 20) ISFSI .........................
...............
................................... Independent Spe nt Fuel Storage Installation Keff .........................
..................................
......................... Effective Neutron Mu ltipli cation Fac t or LCO ..........................
.........................
........................................
.... Lim itin g Co ndition of Operation
.......................................................................................................................................... L OCA .....................
..................................................................................... Loss of Coo l ant Acci d e nt ..........................................................................
................................................................ r,.,4 CR .................................................
............
...........................................
............. +/-1,{a ia Control Roo1n ...........
............................................................................................................................... r,.,4 SIV .........................................................
...........................
....................
+/-1,<laia Steam Isolation Valve +/-1,{SL .......................
.......................................................
......................................... +/-1,{ain Steam Line mR , mRem , mrem , mREM ............................................................ milli-Ro e ntgen E quivalent Man MW .....................................................................................................
............................... Megawatt NEI .........................................
...................................
.................................
Nuclear E n ergy Institute
............................
.............................................................................................................. l'J PP .......................................
...........................................
................................... l'l"uolear Po\ver Plaat ...................................................................................................
....................................... N RC .............................
..........................
................................
.......... Nuclear Regulatory Commission l'l"SSS ..................................................
............................................... l'l"uclear Steam Supply Syste1n .................................................................................................................
......................... N ORAD ................
....................................................
North American Aerospace Defense Command (+/-'JO)UE .........................................................
.................................
(+/-'Jotificatioa Of) Unusual Eve nt NUMARC 1 *************************************************************** N ucl ear Management and Resource s Co uncil OBE .............
.......................................................
................................... Operating Basis Earthq u ake OCA .................................................................
............................................ Owne r Contro ll ed Area .......................................................................................................................................... 0 DC M'ODAM ......................................................... Off s ite Dose Calc ul ation (Assessmeat)
Man ual ORO ...........................
......................................
............................... Off s ite Response Organization PA .............................................................................................................................. Protected Area .......................................................................................................................................... P ACS ............................................
..........................................
Priority Aotuation and Control S)'Stem PAG ............................................................
.........................
.................. Prot ective Action Guideline
.......................................................................................................................................... P ICS ................................................................................... Process Information and Control System PRA/PS A .................................... Probabilistic Risk Assessment
/ Probabili s tic Safety Assessment PWR ..................
................................................
...................................... Pressurized Water Reactor ****************************************************************************************************************************************** p S ............................
............................................................................................... Protection System PSIG .........................
...........................................
............................. Pounds per Sq u are Inch Ga u ge R ..............................
........................................................................................................... Roen t gen 1 NUMARC was a pr e d ecessor o rgani za tion of th e N u c l ear E ner gy In s titut e (NE!). 
}lei 99 01 (Revi sion 6) }'Jovelf!ser 2012 ................................................
.................
......................................................................... R CC .............................................................................................................. Reactor Control Console .....................................................................................................................
..................... R CIC ...................
....................................................
........................... Reactor Core Iso l ation Cool i ng R CS .....................
.........................
.......................................... .' .................... Reac t o r Coolant Syste m R e m , rem , REM ...................................................................................... Roentge n Equivalent Ma n .................................................................
......................................................
...................
R ETS ..............................................
...........................
Radiological Effluent Techn i cal Specifications RPS ......................................................................................................... Reactor Protection System RPV .............................................................
................................................ Reactor Pressure Vessel ............
.............................
..........................................................
...........
............................
R VLIS ......................................................................... Reactor Vessel Level Inst rum e n tat i on System ..........................................................
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..................................... R WCU ............................................................................................................ Reactor \Vater Cleanup ..........................................
................................................................................................ s AR .......................................................................................
......................... Safety i\na l ysis Report ***************************************
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s l\S .................................................
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................ Safet)* ,i\u tomation S)*s tem SBO ......................................................................................................................... Station Blackout SCBA .................
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... Se l f-Co nt ained Breathing Apparatus SG .......................................................................................................................
.... Stea m Ge n e rat or SI ....................................................
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....................................................... Safety Inject i o n ****************************************************************************************************************************************** s ICS ....................................................................
................. Safety Information and Control System **********************************************************************
******************************************************************** s PDS ...........................................................................
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........ Safety Parameter Displa y System S RO ............................................................................................................ Senior Reactor Operator TEDE ............................................................................................. Tota l Effective Dose Equiva l e n t .......................................
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.................................................................... T OJ\F ..................................................................................................................... Top of Active Fuel TSC .......................................................................................................... Tech nic a l S upp ort Center UE ................................................
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U n usual Event UFSAR ..........................................................................
....... Updated F i nal Safety Analys i s Report WOG .................................................................................................. Westingho u se Owners Group l>lel 99 O I (Re\*isieA 6) l>levemaer 2012 APPENDIX B -DEFINITIONS
~ml 99 O 1 (Re~*isieA
: 6) ~Jeyeffiber 2012 The following definitions are taken from Title 10 , Code of Federal Regulations, and related regulatory guidance documents.
Alert: Events are in progress or have occurred which involve an actua l or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels. General Emergency:
Events are in progress or have occurred which involve actua l or IMMINENT substantial core degradation or melting w ith potential for loss of containment integrity or HOSTILE ACTION that results in an actual l oss of physical contro l of the faci lit y. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Notification of Unusual Event (}l"OUEUE/: Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systemsSAFETY SYSTEMS occurs. Site Area Emergency:
Events are in progress or have occurred which involve actua l or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that cou ld lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. _Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary. The following are key terms necessary for overall understanding the NEI 99-01 emergency classification scheme. Emergency Act ion Leve l (EAL): A pre-determined, site-specific, observable threshold for an Initiating Cond ition that , when met or exceeded , places the plant in a g i ven emergency classification level. Emergency Class ifi cation Level (ECL): One of a set of names or tit l es established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences , and (2) resulting onsite and offsite response actions. The emergency classification l eve l s , in ascending order of severity , are: * :Notification of Unusual Event (NG UE)
* A l ert
* Site Area Emergency (SAE) I + This terA1 is seffieti111es sherleAea te UAt1st1al E:~*e11t (UE;) er ether siA'lilar site Sf)eeifie terA'liAelegy.
B-1
* General Emergency (GE) B-2 1>JBI 99 Ql (RevisieR
: 6) 1>le~*emeer 2Q 1 2 Ne! 99 0 1 (R ev i s i o n 6) No v eFRe er 2 0 12 Fission Product Barri er Threshold:
A pre-determined , site-specific , observable threshold indicating the loss or potentia l loss of a fission product barrier. Initi at ing Condition (IC): An event or condition that aligns with the definition of one of t h e four emergency classification l eve l s by virtue of the potential or actual effects or consequences.
Selected terms used in Initiating Condition and Eme rgency Action Level statements are set in all capital letters ( e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document.
The definitions of these terms are provided below. CONFINEMENT BOUNDARY: (In s ert a s ite specific definition for thi s term.) Developer Note -The barrier(s) b etween spent fuel and the environment once the spent fuel is processed for dry storage. CONTAINMENT CLOSURE: (lA s ert a site specific definition for this term.) De>,reloper Note The procedurally defined conditions or actions taken to secure containment (primary or secondary for B'.l/R) and its associated str u ctures, systems, and components as a functional barrier to fission product release under sh utd own conditions.
EXPLOSION: A rapid , vio l ent and catastrophic fai lure of a piece of equipment due to combustion , chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrica l component failure (caused by short circuits , grounding, arcing, etc.) should not automatically be cons id ered an exp l osion. Such events may require a post-event inspection to determine if the attributes of an explosion are present. FAULTED: T he term applied to a steam generator that has a steam leak on the secondary side of s u fficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become comp l etely depressurized. Developer Note This term is applicable to PWR s on l y. FIRE: Combust ion characterized by heat and light. Sources of smoke such as slipping d r ive belts or overheated electrica l equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTIL E ACTION: An act toward a NPPnuclear power plant or its personnel that includes the use of violent force to destroy equipment , take HOSTAGES , and/or intim idate the licen see to achieve an end._ This includes attack by air, l and, or water using guns , explosives , PROJECTILEs , vehicles, or other devices u sed to deliver destructive force. _Other acts that satisfy the overa ll intent may be included.
HOSTILE ACTION should not be construed to inclu d e acts of civil disobedience or felon i o u s acts that are not part of a concerted attack on the NPPnuclear power plant._ Non-t errorism-based EALs should be used to address such activities (i.e.,_-this may include v i o l ent acts between individua l s in the owner controlled area). B-3 
]>JE;I 99 0 I (Re\*i s ieA 6) l>levember 2012 HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception , equipped with suitable weapons capable of killing , maiming , or causing destruction.
IMMINENT:
The trajectory of events or conditions is s uch that an EAL will be met within a re l atively short period of time regardless of mitigation or corrective actions. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of s pent nuclear fuel and other radioactive materials associated with spent fuel storage. NOID,,4:AL LEVELS: As applied to radiological lC/EALs , the highest reading in the past twenty four hours e,wluding the current peak value. OWNER CONTROLLED AREA: (Insert a site specific definition for this terrn.) Devel013eF Nete This term is typically taken to mean the site property owned by , or otherwise under the control o f, the licensee. In some cases , it may be appropriate for a licensee to define a smaller area with a perimeter closer to the plant Protected Area perirneter (e.g., a site 1 , 1 t'ith a large OCA where some portions of the boundary rnay be a significant distance from the Protected Area). In these cases , developers should consider using the boundary defined by the Restricted or Secured Owner Controlled Area (ROCNSOCA). The area and boundary selected for scheme use must be consistent with the description of the same area and boundary contained in the Security Plan. PROJECTILE:
An object directed toward a NWnu c le a r po we r plant that could cause concern for its continued operability , reliability , or personnel safety. PROTECTED AREA:_ (In s ert a site specific definition for this terrn.) Denl013eF Nete This terrn is typically taken to mean ti he area under continuous access monitoring and control , and armed protection as described in the site Security Plan. REFUELING PATHWAY:_ (Insert a site specific definition for this term.) DeYel013eF Nete This description should in c lude all the cavities , tube s, canals and pools through *,vhich irradiated fuel rnay be moYed, but not including the reactor Yessel.The reactor refueling cavity. spent fuel pool and fuel transfer canal. RUPTURE(D):
The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
Devel013eF Nete This terrn is applicable to PWRs only. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems are classified as safety-related.
Devel013eF Nete This term may be modified to include the attributes of " safety related" in accordance with 10 CFR 50.2 or other site s pecific terminology, if desired. B-4 
}IE I 99 0 1 (R e vi s i o A 6) NoveA'!b er '1 0 12 SECURITY CONDITION:
Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. _A SECURITY CONDITION does not involve a HOSTILE ACTION. SJT E BOUNDARY:
T hat lin e bevond which the l a nd is ne i ther owned. nor l ea s e d. nor oth e r w ise controll ed b y the li ce n see. UNISOLABLE:
An open or breached system line that cannot be isolated , remotely or locally. UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements , testing , or analysis.
_The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Damage resulting from an equipment failure and limited to the failed component (i.e .. the failure did not cause damage to a structure or any other equipment) is not VISIBL E D A M A G E. B-5 l>lel 99 0 I (Re\1 i s i0A 6) l>Jevemeer 2012 APPENDIX C PERMANENTLY DEFUElED STATION ICs/EAls Recognition Category PD provides a stand alone set ofICs/EALs for a Permanently Defueled nuclear pov1er plant to consider for use in developing a site specific emergency classification scheme. For development , it 1 ,Yas assumed that the plant had operated under a 10 CFR &sect; 50 license and that the operating company has permanently ceased plant operations.
Further , the company intends to store the spent fuel within the plant for some period of time. When in a permanently defueled condition , the plant licensee typically receives approval from the NRG for eJcemption from specific emergenCJ' planning requirements.
These e1cemptions reflect the lov, , ered radiological source term and risks associated with spent fuel pool storage relative to reactor at power operation. Source terms and accident analyse s associated with plausible accidents are documented in the station's Final Safet)' Analysis Report (FSAR), as updated. As a result , each licensee will need to develop a site specific emergency cla ss ification scheme us i ng the NRG approved eJ1.emptions , revised source terms , and re*,ised accident analyses as documented in the station's FSAR. Recognition Category PD uses the same ECLs as operating reactors; howeYer , the source term and accident analyses typically limit the ECLs to an Unusual Event and Alert. The Unusual Event ICs provide for an increased awareness of abnormal conditions v , hile the Alert ICs are specific to actual or potential impacts to spent fuel. The source terms and release motive forces associated 1 Nith a permanently defueled plant 1 Nould not be sufficient to require declaration of a Site ,<\rea Emergency or General EmergenCJ'. , 11 , permanently defueled station is essentially a spent fuel storage facility vl'ith the spent fuel is stored in a pool of water that serves as both a cooling medium (i.e., removal of decay heat) and shield from direct radiation.
These primary functions of the spent fuel storage pool are the focHs of the Recognition Category PD ICs and EALs. Radiological efffoent IC and EALs were inclHded to provide a basis for clas s ifying events that cannot be readily classified based on an observable events or plant conditions alone. Appropriate ICs and EALs from Recognition Categories A , C , F , H , and S 1 Nere modified and included in Recognition Category PD to address a spectrum ofthe events that may affect a spent fuel pool. The Recognition Category PD ICs and EALs reflect the relevant gHidance in Section 3 of this document (e.g., the importance of avoiding both over classification and under classification). : Nonetheless , each licensee will need to develop their emergencJ' classification scheme using the NRG approved eJcemptions , and the source terms and accident analyses specific to the licensee.
Secu r ity related events 'Nill also need to be considered. C-1 
}Jel 99 01 (Re\*isieA 6) }Jayemser 2012 Table PD 1: Reeognition Category "PD" Initiating Condition Matrix UNUSUAL EVENT PD AUl Release of gaseous or liquid radioactivity greater than 2 times the (site specific effluent release controlling document) limits for 60 minutes or longer. Op. Mades: , Vat AJJJJlietlbk PD AU2 UNPLA1'il'rED rise in plant radiation levels. Op. },fades: Nat AJJJJlieeble PD SUl illtPLA1'il'rED spent fuel pool temperature rise. Op. },fades: ]'latAJJJJlieebk PD HUl Confirmed SECURITY C01'IDITI01'J or threat. Op. },fades: ,V8t AJJJJlieebk PD HU2 Hazardous event affecting SAFETY SYSTEM equipment necessary for spent fuel cooling. Op. Modes: Nat AJJJJlieebk PD HUJ Other conditions e,dst which in the judgment of the Emergency Director warrant declaration of a (1'JO)UE.
Op. },fades: ,V8t AJJJJlieabk ALERT PD .AA.1 Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Op. },fades: ,V8t AJJJJlieebk PD AA2 UN:PLAN1'IBD rise in plant radiation levels that impedes plant access required to maintain spent fuel integrity.
Op. },fades: l 1 latAJJJJlieeblc PD WA HOSTILE ACTION 1tvithin the O\V1'JER C01'ffROLLED AREA or airborne attack threat within 30 minutes. Op. },fades: ]'lat AJJJJlieable PDHA3 Other conditions e,tist *n<h i ch in the *udgment of the Emergency Director ,.,*arrant declaration of an Alert. Op. },fades: NatApplieeble I Table intendea for use by 1 EAL Ele&#xa5;elopers.
: Inclusion i:n licensee C-2 I ,J * * ,J 1 uocuments 1s not reqmreu. 1 L------------------J ECL: Notification of Unusual Event }l!;I 99 0 I (Revi s i o A 6) }/ovem e er 2012 PD AU1 Ieitiatieg Ceeditiee:
Release of gaseous or liquid radioactivit)
* greater than 2 times the (site specific effluent release controlling document) limits for 60 minutes or longer. Netes: Of)erating Mede Af)f)lieability:
Not Applicable Examf)le Emergeeey Aetiee Le1,*els: (I or 2)
* The EmergencJ' Director sJ:wuld declare the Unusual Event promptly upon determining that 60 minutes has be e n e1..ceeded , or will likely be e1weeded.
* If an ongoing release is detected and the release s tart time i s unknown , assume that the release duration has e1tceeded 60 minutes.
* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path , then the effluent 1'l'lonitor reading is no longer valid for classification purposes. (1) (2) Reading on A..""*&#xa5; effluent radiation monitor greater than 2 times the alarm setpoint established bJ' a current radioa c tivity discharge permit for 60 minutes or longer. Sample analysis for a gaseou s or liquid release indicates a concentration or release rate greater than 2 times the (site specific effluent release controlling document) limits for 60 minutes or longer. Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low level radiologica l release that e 1 weeds regulatory commitments for an eJttended period of time (e.g., an uncontro ll ed release).
It includes any gaseous or liquid radiological release , monitored or un monitored , including those for *.vhich a radioactivity discharge permit is normally prepared. Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment.
Further , there are administrative controls established to prevent unintentional release s, and to control and monitor intentional relea s es. The occurrence of an eJttended , uncontrolled radioacfr,e release to the environment i s indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifY i ng events and conditions that cannot be readily or appropriately clas s ified on the basis of plant conditions alone. The inclus i on of both plant condition and radiological effluent E ALs more fu l ly addresses the spectrum of possible accident event s and conditions.
C-1 l>JEI 99 0 I (ReYisieA
: 6) NeYefftaer
?012 Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Releases should not be prorated or averaged.
For example, a release e>weeding 4 times release limits for 30 minutes does not meet the EAL. EAL #1 This EAL addresses radioactivity releases that cause effluent radiation monitor readings to e>weed 2 times the limit established by a radioactivity discharge permit. This EAL will ty*pically be associated with planned batch releases from non continuous release pathv,1ays (e.g., radwaste, v,aste gas). EAL #2 This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample ana(;,ses or environmental surveys, particularly on unmonitored pathv,ca;'s (e.g., spills of radioactive liquids into storm drains, heat e>whanger leakage in river *water systems , etc.). Escalation of the emergency classification level would be via IC PD AAl. Develaf)eF Nates: The "site specific effluent release controlling document" is the Radiological Effluent Technical SpeeifieatioRs (RETS) or , for plants that have implemented Generic Letter 89 01 ++,. the Offsite Dose Calculation Manual (ODCM). These documents implement regulations related to effluent controls (e.g., 10 CFR Part 20 and 10 CFR Part 50, Appendbt l). As appropriate, the RETS or ODCM methodology should be used for establishing the monitor thresholds for this IC. Listed monitors should include the effluent monitors described in the RETS or ODCM. Developers may also consider including installed moRitors associated with other potential effluent pathways that are not described in the RETS or ODCMHH_ If included, EAL values for these monitors should be determined using the most applicable dose/release limits presented in the RETS or ODCM. It is recognized that a calculated EAL value may be belov, what the monitor can read; iR that ease, the monitor does not need to be iReluded in the list. Also , some monitors may not be governed by Technical Specifications or other license related related requirements; therefore, it is important that the associated EAL and basis section clearly identify any limitations on the use or availabilit;
' of these monitors.
++ !m19!ementc1/ien
&j Pregr*lm111*ltie Ce,*1trels f<,r R*ld-ie!egie*l!
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8j9eeijieatie11s in the Adminis,'rc1th ,e C8ntr8fs &et/811 of the Teeh1'lie*l!
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13laA sectieA(s)
*,*,hich aEIElress the req1:JiremeAts ef IO CFR 50.47(13)(8) aAEI (9). +.. DeYele13ers shettlEI kee13 iA miAEI the req1:JireA~eAts ef IO CFR 50.54(q) aAEI the g1:JieaAce 13re~*ieeEI ay fNPO relates te emergeAC)'
res13eAse eq1:Ji13FAeAt wheA ceAsiEleriAg the aEIElitieA efether eftll:leAt FAeAiters.
C-2 J>Je l 99 O I (R e Yi s ieR 6) Ne v e mber 2012 Some sites may find it advantageous to address gaseous and liquid releases v,rith separate Radiation monitor readings should reflect values that correspond to a radiological release e~weeding 2 times a release control limit. The controlling document typically describes methodologies for determining effluent radiation monitor setpoints; these methodologies should be used to determine EAL values. In c a s es where a methodology i s not adequately defined , developer s should determin e value s con s i s tent with effluent control regulations (e.g., 10 CFR Part 20 and 10 CFR Part 50 Appendi><, I) and related guidance.
For E i\L #1 Values in this EAL should be 2 times the setpoint established by the radioactivity discharge permit to warn of a release that i s not in compliance with the specified limits. Indexing the value in thi s mann e r en s ures consi s ten c y between the E AL and the s etpoint established by a specific discharge permit. Developers should research radiation monitor design documents or other information sources to ensure that 1) the EAL value being considered is withiH the usable response and display range of the instrument , and 2) there are no automatic features that may render the monitor reading invalid (e.g., an auto pur g e feature triggered at a particular indication level). It is recognized that the condition described by this IC may result in a radiological effluent value bey'ond the operating or di s play range of the installed effluent monitor. In those cases , EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available.
For e M.ample , an E AL monitor reading might be set at 90% to 95% of the highest accurate monitor readin g. Thi s provision notvt'ithstanding , if the estimatedJcalculated monitor reading i s greater than approximately 110% of the highest accurate monitor reading, then developers ma y choo s e not to include the monitor as an indication and identify an alternate BAL threshold. Indications from a real time do se projection sy s tem are not included in the generic E ALs. Many licensees do not have this capability.
For those that do , the capability may not be within the scope of the plant Technical Spe c ifi c ation s. A licensee may request to include an EAL using real time dose projection system result s; approval 1 ,Yill be considered on a ca s e by case basis. Indications from a perimeter monitoring system are not included in the generic EALs. Many licensees do not hav e thi s capabilit y. For those that do , these monitors may not be c ontrolled and maintained to the same level a s plant equipment , or within the scope of the plant Technical Specifications. In addition , readings may be influenced by environmental or other factors. A licensee may request to include an EAL using a perimeter monitoring system; approval will be considered on a case b y ca s e ba s is. EG L Assignment Attributes:
3.1.1.B C-3 
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(~ H0!5!A0"8) 10 66 (:I N Develeper Netes: 1>11;1 99 0 I (Re,*i s ien 6) l>J 0Y0ffi00r 20 J 2 For EAL # 1 Site specific indications may include instrumentation 11alues such as water level and area radiation monitor readings , and personnel reports. If available, 11ideo cameras may allow for remote observation. Depending on available instrumentation , the declaration may also be based on indications of 'tvater makeup rate and/or decreases in the le11el ofa water storage tafHt For EAL #2 The specified
*,alue of 25 mR/hr may be set to another 11alue for a specific application with appropriate justification.
EGL Assignment Attributes:
: 3. l. l .B l>JE I 99 01 (R ev i s i oR 6) l>/0\'0ffi0 0F 2 0 1 2 ECL: Notification ofUnusua:l E~*=,e=n~t
_______________
gpo SU1 Initiating Condition: J"PLAl'l" N E D spent fuel pool temperatur e rise Operating Mode A 1* * *
* app iealnhty:
l'fot Applicable Example Emergeney Aetion Lenis: (l) p e e temperature rise to g reater than (s ite specific 0 UNPLANNED spent fuel I I+. This I? addresses a condition that i potential degrad~ion in the level o; s:::~~:;s t:r tel a more s eriou s event and represents a occur , and result ma loss of pool level and . e p ~nt. l!u.ncorre c ted , boiling in the pool will . mcrease rad1at1on levels. Escalation of the emer en. , . . g C) clas s ification.
le" el lJ'ould b . r fr e via IC PD A A D...,lapeF Noles, "" ,I e, PB AA2. lelflpOfft!UfO i, HeRHOI ~-ffi1~**"}~
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is 125" le 15 9~:e~;.re:
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* e site specific temperature should b caleulations in the SAR T , -~ chosen. ba s ed on th e startin . '"g syslem malffiHe!iaH p,ier 10 ~;;;;,;-:~';
*-15 peiHI "'"' allewiHg time le 00;.-.~*!1,e IOn. EGL Assign.ment Attribute s: 3 .1.1.A B-3 ECL: Notification of Unusual Event Initiating Condition:
Confirmed SECURITY CONDITION or threat. Operating Mode i ... pplieability:
Not Applicable Example Emergeney Aetion Lenis: (1 or 2 or 3) J>lE;I 99 0 I (Re\*i s ieA 6) N e vefl'!eer 2012 PD HU1 (1) A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the (site specific security shift supervision). (2) (3) Notification of a credible security threat directed at the site. A Yalidated notification from the NRG proYiding information of an aircraft threat. This IC addresses events that pose a threat to plant personnel or the equipment necessar)' to maintain cooling of spent fuel, and thus represent a potential degradation in the !eye) of plant safety. Security events which do not meet one of these Ei*\Ls are adequately addressed by the requirements of 10 CfR &sect; 73.71 or 10 CfR &sect; 50.72. Security e*,cents assessed as HOSTILE ACTIONS are classifiable under IC PD HAI. Timely aHd accurate cmHmuHications bet\veen Security Shift Super,ision and the Control Room is essential fur proper classification of a security related event. Classification of these events will initiate appropriate threat related notifications to plant personnel and OROs. Securit)1 plans and terminology are based on the guidance provided by NEI 03 12 , Tcmplalefor the Seeitrity Plan , Training and Qualifieatifm Pl-an , Sa.fcguat<ds Centingency Pl-ati 8'-md Indcpendcnl Spent Fite! Sterage Installalifm Sceitrity Program}.
EAL #1 references (site specific security shift supervision) because these are the individuals trained to confirm that a security event is occurring or has occurred.
TraiHing on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CfR &sect; 2.39 information.
EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance
*uith (site specific procedure).
EAL #3 addresses the threat from the impact of an aircraft on the plant. The :NRG Headquarters Operations Officer (HOO) v,ill communicate to the licensee if the threat involves an aircraft.
The status and size of the plane may also be provided by NORAD through the :NRG. Validation of the threat is performed in accordance with (site specific procedure).
Emergency plans and implementing procedures are public documents; therefore , EA.Ls should B-4 l>JJ;I 99 0 I (R e ,*isieA 6) l>J e yemeer 2012 not incorporate Security sensitive information.
This includes information that may be advantageous to a potential adver s ary , such as the particulars concerning a specific threat or threat location.
Security sensitive information should be contained in non public documents such as the Security Piaf!. Escalation of the emergency classification level v , ould be via IC PD HAI. Develaper Netes: The (site specific security s hift supervi s ion) is the title of the on s hift individual responsible for supervision of the on shift security force. The (site specific procedure) is the pro c edure(s) used by Control Room and/or Security personnel to determine if a securit y threat is credible , and to validate receipt of aircraft threat information.
Emergency plans and implementing pro c edures are public documents; therefore , BALs should not incorporate Security sensitive information.
This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
Securit)1 sensitive information should be contained in non public documents such as the Security Plan. With due consideration given to the abov e de*,celoper note , EALs may contain alpha or numbered references to selected events described in the Securit)' Plan and a s sociated implementing procedures. Such reference s should not contain a re c ognizable de s cription of the event. For e~{ample , an EAL may be worded as " Security event #2, #5 or #9 is reported b)1 the (site specifi c security shift supervision)
." E GL Assignment Attribute s: 3. I. I .A 
~le i 99 0 I (Re~*i s ioR 6) ~loYeffiber 2012 PD HU2 ECL: *Notification of Unusual Event Initiating Condition:
Hazardous event affecting SAFETY SYSTEM equipment necessary for spent fuel cooling. Operating Mode AfJJJlieahility:
Not Applicable ExamJJle Emergeney ,A .. etion LeYels: (1) a. b. C. The occurrence ofA .... W of the follovving hazardous events:
* Seisrnic event (earthquake)
* Internal or eJ(ternal flooding event
* High winds or tornado strike
* FIRE
* EXPLOSIOJli~
* (site specific hazards)
* Other events with similar hazard characteristics as deterrniHed by the Shift Manager The eYent has damaged at least one tram of a SAFETY SYSTEM needed fur spent fuel cooling. The damaged SAFETY SYSTEM train(s) cannot , or potentially cannot , perform its design function based on EITHER:
* Indications of degraded performance
* VISIBLE DAMAGE This IC addresses a hazardous eyent that causes damage to at least one train of a SAFETY SYSTEM needed for spent fuel cooling. The damage must be of sufficient rnagnitude that the system(s) train cannot , or potentially cannot, perform its design function.
This condition reduces the rnargin to a loss or potential loss of the fuel clad barrier, and therefore represents a potential degradation of the ]eye] of safety of the plant. For EAL l .c , the first bullet addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. B-6 l>Jl~I 99 01 (R e~*isieA 6) ~le vemser 2012 For EAL l .c , the second bullet addresses damage to a SAFETY SYSTEM train that is not in service/operation or readily apparent through indications alone. Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment n o t requiring lengthy analysis or quantification of the damage. Esealation of the emergency clas s ifi e ation level could, depending upon the event , be based on any of the Alert !Cs; PD AA I , PD AA2 , PD HA I or PD HA3. DeYelaf)er Nates: For (site specifie hazards), developer s should eonsider including other significant , site specifie hazards to the bulleted list contained in E AL l .a (e.g., a seiche). Nuelear pov,rer plant Si"5ETY SY8 TE M 8 are comprised oftv10 or more separate and redundant trains of equipment in aeeordance w ith site specific design eriteria.
EGL Assignment Attributes:
3 .1.1.A and 3 .1.1 C B-7 l>J E I 99 01 (RevisieA 6) l>JeYeFASer 20 [ 2 PD HU3 ECL: Notification of Unusual Event Initiating Cenditien:
Other conditions e>,ist which in the judgment of the Emergency Director warrant deelaration of a a,JO)UE. Operating Mede Applieahility:
Not Applicable Example Emergeney Aetien Le*/els: (]) Other conditions e>dst 1 which in the judgment of the Emergency Director indicate that events are in progress or have occurred \'>'hich rndicate a potential degradation of the level of safety of the ploot or indicate a security threat to facility protection has been initiated. Jl,fo releases of radioacfr,e material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. This IC addresses unanticipated conditions not addressed e1(plicitly elsewhere but that 1.varrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a Jl.tOUE. B-8 PD AA1 ECL: Alert leitiatieg Cond .. greater than 10 m itwn: Release of g rem TEDE O aseous or Ii . . 0 . , 50 ..,,..., thyroid CDE quid rad10aotivity res*kia . . peralmg Med , .
* g ** offs , e uppheahility:
N t A
* 0 ' ,pphcable LeYels* El
* or or 3 or 4) Notes: (1) (3) (4)
* Closed window ude:osste~d:,,.
r or longer. rates greater than 1 o R
* Aaalyses eflield ..,~11,, e,peeted to eeatia he*r ef iaholatie.'"""Y Sftfflples iadieate ti, . ** Hl, 60 lfliautes . ry)''ftro:tt1-a-d-{:C;ID}le~:reaiteF-ih:
..
* greater than 50 mrem for one Basis: B-9 l>lel 99 01 (ReYi s ieA 6) l>l 9\'eFAser ?Q ( 2 This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protecfr,e Action Guides (PAGs). It includes both monitored and un monitored releases.
Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated b)' a radiological release that significantly e>weeds regulatory limits (e.g., a significant uncontrolled release).
Radiological effluent EAbs are also ineluded to provide a basis for elassifying events and eonditions that eannot be readily or appropriately classified on the basis of plant eonditions alone. The inelusion of both plant eondition and radiologieal effluent &#xa3;Abs more fully addresses the speetrum of possible aecident events and conditions. The TEDE dose is set at 1% of the &#xa3;PA PAG of 1 , 000 mrem *.vhile the 50 mrem thyroid COE was established in eonsideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid GO&-Classifieation based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flo,v past an effluent monitor is known to have stopped due to actions to isolate the release path , then the effluent monitor reading is no longer valid for classification puff)oses.
Develeper Netes: While this IC may not be met absent challenges to the cooling of spent fuel , it provides classifieation di*,ersity and may be used to classif)' events that ,,,.*ould not reach the same EGL based on plant conditions alone. The &#xa3;PA PAGs are ex.pressed in terms of the sum of the effective dose equivalent (EDE) and the committed effecfr,e dose equivalent (CEDE), or as the thyroid eommitted dose equivalent (COE). For the purpose of these IC/EAbs , the dose quantit)' total effective dose equivalent (TED&#xa3;), as defined in 10 CFR &sect; 20 , is used in lieu of" ... sum of EDE and CEDE .... ". The EP,c\ PAG guidaRce provides for the use adult thyroid dose conversion factors; however, some states have decided to base protective aetions on child thyroid COE. Nuelear power plant ICs/EAbs need to be consistent
't&#xa5;ith the proteetive aetion methodologies employed by the States within their EPZs. The thyroid COE dose used in the IC and &#xa3;Abs should be adjusted as neeessary to align with State protecfr,e action decision making eriteria.
The " site specific monitor list and threshold values" should be determined with consideration of the following:
* SelectioR of the appropriate installed gaseous and liquid effluent monitors.
* The effluent monitor readings should eorrespond to a dose of 10 mrem TEDE or 50 rnrem thyroid COE at the "s ite specific dose receptor point" (eonsistent with the caleulation methodology employed) for one hour of e>(posure.
* Monitor readings will be calculated using a set of assumed meteorological data or B-10 N e l 99 01 (Re,*isieA 6) ~10Y8Ffl08F 2012 atmospheric dispersion factors; the data or factors selected for use should be the same as those employed to calculate the monitor readings for IC PD AUI.
* The calculation of monitor readings 1.vill also require use of an assumed release isotopic miJ{; the selected mb{ should be the same as that employed to calculate monitor readings for IC PD AUL
* Depending upon the methodology used to calculate the El.,L values , there may be overlap of some values between different ICs. Developers 1 ,1t*ill need to address this overlap by adjusting these values in a manner that ensures a logical escalation in the EGL. The " site specific dose receptor point" is the distance(s) and/or locations used by the licensee to distingui s h between on site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan , and the procedural methodology used to determine offsite doses and Protective Action Recommendations.
The variation in selected dose receptor point s means there may be some differences in the distance from the release point to the calculated do se point from site to site. Developers s hould research radiation monitor design documents or other information sources to ensure that I) the EAL Yalue being considered is within the usable response and display range of the instrument , and 2) there are no automatic features that may render the monitor reading invalid (e.g., an auto purge feature triggered at a particular indication le v el). It is recogni'l ed that the condition described by this IC ma;* result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. In those cases , EAL values should be determined 1tYith a margin sufficient to ensure that an accurate monitor reading is available.
For eJc.ample , an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This proYi s ion notwithstanding , if the estimated/calculated monitor reading is greater than apprm{imately 110% of the highest accurate monitor reading , then de1t*elopers may choose not to include the monitor as an indication and identify an alternate E AL threshold.
Although the IC references TEDE , field survey results are generally available only as a " whole body" dose rate. For this reason , the field s1:1rvey EAL specifies a " closed window" survey reading. Indications from a real time dose projection system are not included in the generic EALs. Many licensees do not have this capability.
For those that do , the capability may not be within the scope of the plant Technical Specifications.
A licensee may request to include an EAL using real time dose projection system res1:1lts; approval will be considered on a case by case basis. Indications from a perimeter monitoring system are not included in the generic EALs. Many licensees do not have this capability. For those that do , these monitors may not be controlled and maintained to the same level as plant equipment , or within the scope of the plant Technical Specifications.
In addition , readings may be infll:lenced by environmental or other factors. A licensee may request to include an EAL using a perimeter monitoring system; approval will be considered on a case by case basis. B-11 EGL i\ssignmeAt Attributes:
3.1.2.C B-12 l>JEI 99 0 I (RevisieA
: 6) l>le~*effiaer 2012 l>le l 99 0 I (R e ,*isieR 6) N e&#xa5;eml3er 2012 PDAA2 ECL: Alert Initiating Cenditien:
illl"PLA1'n>nm rise iR plant radiatioR levels that impedes plant access required to maiRtaiR speRt fuel iRtegrity. 013erating Mede A1313lieahility:
Not Applicable (1) (2) Exam13le Emergeney Aetien Levels: (1 or 2) illl"PLAl'H>l"ED dose rate greater thaA 15 mR/hr iR A...~Y of the followi11g areas requiri11g c0Rti11uous occupa11ey to mai11taiR eo11trol of radioactive material or operatioR of systems Reeded to maiRtaiR speRt fuel integrity: (site specific area list) illJPLi\"Nl'l"ED Area RadiatioR Mo11itor readiRgs or survey results iRdicate a rise by 100 mR/hr over l'JOR,\4AL LEVELS that impedes access to A ... of the followiRg areas Reeded to maiRtaiR coRtrol of radioactive material or operatioA of systems Reeded to maintain speRt fuel iRtegrity. (site specific area list) This IC addresses iRereased radiatioR levels that impede Recessary access to area s coRtaiRiRg equipmeRt that must be operated maRually or that requires local moRitoriRg , iR order to maiRtaiR systems Reeded to maiRtaiR speRt fuel iRtegrity. As used here, 'impede' i11cludes hiRderiRg or i11terferi11g , provided that the iRterfereRce or delay is sufficieRt to sigRificaRtly threateR Reeessary plaRt aeeess. It is this impaired access that results iR the actual or poteRtial substantial degradatioR of the leYel of safety of the plaRt. This IC does Rot apply to aRticipated temporary iRcreases due to planRed eveRts. Devel013er Nates: The value of l 5mR/hr is defrred from the GOG 19 value of 5 rem iR 30 days 1 with adjustment for expected occupaRcy times. A l though SectioR III.D.3 of l'JUREG 0737 , Ckirificetien
<:>}Tl,llActifJli Pkli9 Rcqitircmcnts , provides that the 15 mR/hr va l ue eaR be averaged over the 30 days , the value is used here *without averagiRg , as a 30 day duratioA implies a R eveRt poteRtially more sig11ificaRt than aR Alert. The specified value of 100 mR/hr may be set to aRother value for a specific applicatioR with appropriate justificatioR. EGL AssigAmeAt Attributes:
3 .1.2.C B-13 
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)(OtlUtJ ijlUOJ!l3 Ul3 JO UO!ltJO!.J!lOU JO VffifV 03110"M+/-tfil3
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~'V'H Cd c I oc .1 a~1,1,1 a A 01<t ("3 ll S!S!,~a zi) IO 6 6 !'3t<I:
l>le l 99 O I (Revi s ieH 6) Nevemeer 2012 IY:~ #2. add~e~~,:s ~he thre~t from the i1!1pact of aA. aircraft oA the p laAt , aAd the aAticipated am: al t1~e 1s "1th1A 30 mrnutes. The mteAt ofth1s EAL is to eAsure that threat related Aot1ficat10As.
are made iA a timely maAAer so that plant personnel aAd OROs are in a heightened state ofread1~ess
.. This EAL is met when the threat related information has beeA validated in accordaAce with (site specific procedure).
!he NRG He?dquarter s Operations Officer (HOO) *will communicate to the IiceAsee if the threat ='es an aircraft. The status aAd size of the plane may be provided by : NORAD through the IA some cases , it may not be readily appareAt if an aircraft impact within the O\VNER CONTR?LLED A~A ~.vas iAteAtional (i .. e., a HOSTIL E ACTIO}~). It is e~{pected , although Rot c.ertam , that Aot1fica!10n by an appropF1ate Federal agency to the site would clarify th i s poiAt In this case , the app.ropr~ate federal agency is iAteAded to be NORAD , FBI , FA.,<\ or NRG. The em~rgeAc):
?eclara~10A , .rncluding one based OR other IC s/E ALs , should Rot be uAduly de l ayed ',1t'h1le awa1trng Aot1ficat10A by a Federal ageAC)'. Em:rgeAcy plaAs aAd.implem.e~tin
.g proced.ures are publi c documents; therefore , EALs should no~, rncorporate 8ecuflt)' ~ens1t1ve mformat10A.
This includes iAformatioA that may be ad~ aAtageo~s to a pot:At1al adversary , such as the particular s coAcerning a specific threat or threat locat10A. s.ecuflt)' seA s itive informatioA should be coAtained in non public documeAts such as the Secuflty Plan. DeYelaper Nates: The (si.t~ specific securit)' shift supervision) is the title of the oA s hift iAdividual respoAsible for superv1sI0A of the on shift security force. t s Em:rgency plaA s aAd implementing procedures are public documents* therefore EAL s sh?,uld not mcorporate S:curity sensitive information.
Thi s includes inform~ioA that ~ay be ad, antageo~s to a potential adversary , such as the particulars concerniAg a specific threa t or hreat location. s.ecurity sensitive inforrnatioA should be contained in non public documents uch as the Secuflt)'
Plan. n r thd *d *
* vr1ue COR S I erat10A given to the above developer note, E ALs may coAtaiA alpha or I e ( numbered ~eferences to selected eveAts described iA the 8ecurit)' Plan and associated
~plement1A~
procedures
.6 Such references should Rot contain a recognizable descriptioA of the *. ent. Fo_r e>campl~, aA E , .L may be worded as " Security e't<ent #2 , #5 or #9 is reported b y the site specific secunty s hift s upervision)
." s a ee the related ?e','eloper
}fote in Appendb, B , DefiAitions , for guidance oA the developmeAt of scheme defimt10n for the OWNER CONTROLLED AREA. EGL Assignment Attributes:
3.1.2.D B-15 T>J E I 99 01 (Re~*isieA 6) T>J eyeffieer 2012 PD HA3 ECL: Alert Initiating Ceeditien:
Other condition s eJ (i s t which in the judgment of the Emergency Director ',&#xa5;arrant declaration of an Alert. Operating Mede Applieability:
Not Applicable Example Emergeney i~etien Levels: (1) Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred 1.Yhich involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are e 1 tpected to be limited to small fractions of the EPA ProtectiYe Action Guideline e1(posure le 1 1els. This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions e1dst which are belie 1 ,ied by the Emergency Director to fall under the emergency classification level description for an Alert. B-16 ATTACHMENT 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 286, ADOPTION OF EMERGENCY ACTION LEVEL SCHEME PURSUANT TO NEI 99-01 REVISION 6, "DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR NON-PASSIVE REACTORS" UPDATED CLEAN COPY OF NEI 99-01 REVISION 6 136 pages follow Point Beach Emergency Action Levels Bases Document**
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11 TABLE OF CONTENTS . . 1 . REGULATORY BACKGROUND
.................................................................................................
1 1.1 OPERATING REACTORS ***********************
... ........................
-.......................... ** .._ ******* * ................
1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) ......................................................
2 1.3 NRC ORDER EA-12-051
.............................. ....**....***..................................................*.......
3 2 KEY TERMINOLOGY USED -IN NEI 99*01 ................................................................................
4 2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) ...***...........................................................*..........*
4 2.2 INITIATING CONDITION (IC) ............................................................... ...................................
6 2.3 EMERGENCY ACTION LEVEL *(EAL) ...............*...........*.....*.......................................................
6 2.4 FISSION PRODUCT BARRIER THRESHOLD
****************************************--***--*********
.........................
6 3 DESIGN OF THE PBNP EMERGENCY CLASSIFICATION SCHEME ******************************************
7 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLs) ...................................................
7 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS********--*******************************
11 3.3 PBNP-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION
.........................
12 3.4 IC AND EAL MODE APPLICABILITY
***************************************************************************************
13 4 PBNP SCHEME DEVELOPMENT
............................................................................................
14 4.1 GENERAL DEVELOPMENT PROCESS **************************--****--****--***--******************************************
14 4.2 CRITICAL CHARACTERISTICS
................................................................................................
14 4.3 INSTRUMENTATION USED FOR EALs .....................................................................................
15 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA
************************************
15 5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS
***************************************************
16 5.1 GENERAL CONSIDERATIONS*************************************************************************************************
16 5.2 CLASSIFICATION METHODOLOGY
******************************************************************************************18
 
===5.3 CLASSIFICATION===
 
OF MULTIPLE EVENTS AND CONDITIONS
***************
..........................................
18 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION
**************************************************
18 5.5 CLASSIFICATION OF IMMINENT CONDITIONS
*************************************************--**
......................
19 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING
........................................
19 5. 7 CLASSIFICATION OF SHORT-LIVED EVENTS.****************************************************************************20
 
===5.8 CLASSIFICATION===
 
OF TRANSIENT CONDITIONS
..........................................................................
20 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION
.....................................
21 5.10 RETRACTION OF AN EMERGENCY DECLARATION
............................................................ , *********
21 6 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS ***************************************
22 7 COLD SHUTDOWN/
REFUELING SYSTEM MALFUNCTION ICS/EALS ..................................
39 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ****** *******************
60 9 FISSION PRODUCT BARRIER ICS/EALS ................................................................................
62 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS .....................
78 11
* SYSTEM MALFUNCTION ICS/EALS .....................................................................................
102 APPENDIX A -ACRONYMS AND ABBREVIATIONS
.....................................................................
A-1 APPENDIX B -DEFINITIONS
........................................................................................................
B-1 111 POINT BEACH EMERGENCY ACTION LEVELS BASIS DOCUMENT 1 REGULATORY BACKGROUND
 
===1.1 OPERATING===
 
REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities.
Several of these regulations govern various aspects of an emergency classification scheme. A review of the relevant sections listed below will aid the reader in understanding the key terminology provided in Section 3.0 of this document.
* 10 CFR &sect; 50.47(a)(l)(i)
*
* 10 CFR &sect; 50.47(b)(4)
* 10 CFR &sect; 50.54(q) *" 10 CFR &sect; 50.72(a)
* 10 CFR &sect; 50, Appendix E, IV.B, Assessment Actions
* 10 CFR &sect; 50, Appendix E, IV.C, Activation of Emergency Organization The above regulations are supplemented by various regulatory guidance documents.
Three documents of particular relevance to NEI 99-01 are:
* NUREG-0654/FEMA-REP-l, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]
* NUREG-1022, Event Reporting Guidelines JO CFR &sect; 50. 72 and&sect; 50.73
* Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors 1 
, 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) Selected guidance in NEI 99-01 is applicable to licensees electing to use their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are *consistent with the requirements of 10 CFR 50 and the guidance in NUREG 0654/FEMA-REP-1.
The initiating conditions germane to a 10 CFR 72.32 emergency plan (as described in NUREG-1567) are subsumed within the classification scheme for a 10 CFR 50.47 emergency plan. The generic ICs and EALs for an ISFSI are presented in Section 8, ISFSI ICs/EALs.
IC EUl covers the spectrum of credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to installations or facilities that may process and/or repackage spent fuel (e.g., a Monitored Retrievable Storage Facility or an ISFSI at a spent fuel processing facility).
In addition, appropriate aspects ofIC HUI and IC HAI should also be included to address a HOSTILE ACTION directed against an ISFSI. The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, Ajj_egulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees.
NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent.
Regarding the above information, the expectations for an offsite response to an Alert classified under a 10 CFR 72.32 emergency plan are generally consistent with those for an Unusual Event in a 10 CFR 50.47 emergency plan (e.g., to provide assistance if requested).
Also, the licensee's Emergency Response Organization (ERO) required for 10 CFR 72.32 emergency plan is different than that prescribed for a 10 CFR 50.47 emergency plan ( e.g., no emergency technical support function).
2 1.3 NRC ORDER EA-12-051 The Fukushima Daiic~i accident of March 11, 2012, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, and ultimately led to core damage in three reactors.
While the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling. Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii).
Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license. NRC Order EA-12,.051 states, in part, "All licensees
... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level* conditions by trained personnel:
(1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." To this end, all licensees must provide:
* A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;
* A display in an area accessible following a severe event; and
* Independent electrical power to each instrument channel and provide an alternate remote power connection capability.
NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, " provides guidance for complying with NRC Order EA-12-051.
NEI 99-01, Revfaion 6, includes three EALs that reflect the availability of the enhanced *spent fuel pool level instrumentation associated with NRC Order EA-12-051.
These EALs are included within ICs RA2, RS2, and RG2. 3 2 KEY TERMINOLOGY USED IN NEI 99-01 There are several key terms that appear throughout the EAL methodology.
These terms are introduced in this section to support understanding of subsequent material.
As an aid to the reader, the following table is provided as an overview to illustrate the relationship of the terms to each other. Emergency Classification Level Unusual Event I Alert I sAE I GE Initiating Condition Initiating Condition Initiating Condition Initiating Condition Emergency Action Emergency Action Emergency Action Emergency Action Level (1) Level (1) Level (1) Level (1) * . Operating Mode
* Operating Mode
* Operating Mode
* Operating Mode Applicability Applicability Applicability Applicability
* Notes
* Notes
* Notes
* Notes
* Basis
* Basis
* Basis
* Basis (1) -When making an emergency classification, the Emergency Director must consider all ,, information having a bearing on the proper assessment of an Initiating Condition.
This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information.
In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL. 2.1 EMERG~NCY CLASSIFICATION LEVEL (ECL) One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:
* Unusual Event (UE)
* Alert
* Site Area Emergency (SAE)
* General Emergency (GE) 2.1. l Unusual Event (UE) Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs, Purpose: The purpose of this classification is to assure that the first step in future response has been carried out, to bring the operations staff to a state ofreadiness, and to provide systematic handling of unusual event information and decision-making.
4 
 
====2.1.2 Alert====
Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves _probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG
* exposure levels. Purpose: The purpose of this classification is to assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring if required, and provide offsite authorit1.es current information on plant status and parameters.
2.1.3 Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personn1;:l or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA P AG exposure levels beyond the site boundary.
Purpose: The purpose of the Site Area Emergency declaration is to assure that emergency response centers are staffed, to assure that monitoring teams are dispatched, to assure that personnel required for evacuation of near-site areas are at duty stations if the situation becomes more ser1ous, to provide consultation with offsite authorities, and to provide updates to the public through government authorities.
 
====2.1.4 General====
Emergency (GE) Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA P AG exposure levels offsite for more than the immediate site area. Purpose: The purpose of the General Emergency declaration is to initiate predetermined protective actions for the public, to provide continuous assessment of information from the licensee and offsite organizational measurements, to initiate additional measures as indicated by actual or potential releases, to provide consultation with offsite authorities, and to provide updates for the public through government authorities.
5 
 
===2.2 INITIATING===
 
CONDITION (IC) An event or condition that aligns with the definition of one of the four emergency , classification levels by virtue of the potential or actual effects or cqnsequences.
Discussion:
An IC describes an everit or condition, the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed as a continuous, measurable parameter (e.g., RCS leakage), an event (e.g., an earthquake) or the status of one or more fission product barriers ( e.g., loss of the RCS barrier).
Appendix 1 of NUREG-0654 does not contain example Emergency Action Levels (EALs) for each ECL, but rather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency, or events that could lead to a radiological . emergency, has occurred).
NUREG-0654 states that the Initiating Conditions form the basis for establishment by a licensee of the specific plant instrumentation readings (as applicable) which, if exceeded, would initiate the emergency classification.
Thus, it is the specific instrument readings that would be the EALs. 2.3 EMERGENCY ACTION LEVEL (EAL) A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Discussion:
EAL statements may utilize a variety of criteria including instrument readings and status indications; observable events; results of calculations and analyses;
' entry into particular procedures; and the occurrence of natural phenomena.
 
===2.4 FISSION===
PRODUCT BARRIER THRESHOLD A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Discussion:
Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment.
This concept 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
* Reactor _Coolant System (RCS)
* Containment Upon determination that one or more fission product barrier thresholds have been exceeded, the combination of barrier loss and/or potential loss thresholds is compared to the fission product barrier IC/EAL criteria to determine the appropriate ECL. In some accident sequences, the ICs and EALs presented in the Abnormal Radiation Levels/ Radiological Effluent (R) Recognition Category will be exceeded at the same time, or shortly after, the loss of one or more fission product barriers.
This redundancy is intentional as the former ICs address radioactivity releases that result in certain offsite doses from whatever cause, including events that might not be fully encompassed by fission product barriers ( e.g., spent fuel pool accidents, design containment leakage following a LOCA, etc.). 6 1--. 3 DESIGN OF THE PBNP EMERGENCY CLASSIFICATION SCHEME 3.1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) -An effective emergency classification scheme must incorporate a realistic and accurate assessment of risk, both to plant workers and the public. There are obvious health and safety risks in underestimating the potential or actual threat from an event or condition; however, there are also risks in overestimating the threat as well ( e.g., harm that may
* occur during an evacuation).
The PBNP emergency classification scheme attempts to strike an appropriate balance between reasonably anticipated event or condition consequences, potential accident trajectories, and risk avoidance or minimization.
There are a range of "non-emergency events" reported to the US Nuclear Regulatory Commission (NRC) staff in accordance with the requirements of 10 CFR 50.72. Guidance concerning these reporting requirements, and example events, are provided in NUREG-1022.
Certain events reportable under the provisions of 10 CFR 50.72 may also require the declaration of an emergency.
In order to align each Initiating Conditions (IC) with the appropriate ECL, it was necessary to determine the attributes of each ECL. The goal of this process is to answer the question, "What events or conditions should be placed under each ECL ?" The following sources provided information and context for the development of ECL attributes.
* Assessments of the effects and consequences of different types of events and conditions
* PBNP abnormal and emergency operating procedure setpoints and transition criteria
* PBNP Technical Specification limits and controls
* Offsite Dose Calculation Manual (ODCM) radiological release limits
* Review of selected Updated Final Safety Analysis Report (UFSAR) accident analyses
* Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs)
* NUREG 0654, Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants
* Industry Operating Experience
* Input from PBNP subject matter experts The following ECL attributes were created to aid in the development ofICs and Emergency Action Levels (EALs ). The attributes may be useful in briefing and training settings (e.g., helping an Emergency Director understand why a particular condition is classified as an Alert). 7 3 .1.1 Unusual Event (UE) An Unusual Event, as defined in section 2.1.1, includes but is not limited to an event or condition that involves: (A) A precursor to a more significant event or condition. (B) A minor loss of control of radioactive materials or the ability to control radiation levels within the plant. (C) A consequence otherwise significant enough to warrant notification to local, State and Federal authorities.
 
====3.1.2 Alert====
An Alert, as defined in section 2.1.2, includes but is not limited to an event or condition that involves: (A) A loss or potential loss of either the fuel clad or Reactor Coolant System (RCS) fission product barrier. (B) An event or condition that significantly reduces the margin to a loss or potential loss of the fuel clad or RCS fission product barrier. (C) A significant loss of control of radioactive materials resulting in an inability to control radiation levels within the plant, or a release of radioactive materials to the environment that could result in doses greater than I% of an EPA P AG at or beyond the site boundary. (D) A HOSTILE ACTION occurring within the OWNER CONTROLLED AREA, including those directed at an Independent Spent Fuel Storage Installation (ISFSI). 3 .1.3 Site Area Emergency (SAE) A Site Area Emergency, as defined in section 2.1.3, includes but is not limited to an event or condition that involves: (A)A loss or potential loss of any two fission product barriers -fuel clad, RCS and/or containment. (B) A precursor event or condition that may lead to the loss or potential loss of multiple fission product barriers within a relatively short period of time. Precursor events and conditions of this type include those that challenge the monitoring and/or control of multiple SAFETY SYSTEMS. (C) A release of radioactive materials to the environment that could result in doses greater than I 0% of an EPA P AG at or beyond the site boundary. (D)A HOSTILE ACTION occurring within the plant PROTECTED AREA. 3.1.4 General Emergency (GE) A General Emergency, as defined in section 2.1.4, includes but is not limited to an event 8 or condition that involves: (A) Loss of any two fission product barriers AND loss or potential loss of the third barrier -fuel clad, RCS and/or containment.
* (B) A precursor event or condition that, unmitigated, may lead to a loss of all three fission product barriers.
Precursor events and conditions of this type include those that lead directly to core damage and loss of containment integrity. (C) A release of radioactive materials to the environment that could result in doses greater than an EPA P AG at or beyond the site boundary. (D)A HOSTILE ACTION resulting in the loss of key safety functions (reactivity control, core cooling/RPV water level or RCS heat removal) or damage to spent fuel. 9 3.1.5 Risk-Informed Insights Emergency preparedness is a defense-in-depth measure that is independent of the assessed risk from any particular accident sequence; however, the development of an effective emergency classification scheme can benefit from a review of risk-based assessment results. To that end, the development and assignment of certain ICs and EALs also considered insights from several site-specific probabilistic safety assessments.
Some generic insights from this review included:
-1 .. Accident sequences involving a prolonged loss of all AC power are significant contributors to core damage frequency at many Pressurized Water Reactors (PWRs). For this reason, a loss of all AC power for greater than 15 minutes, with the plant at or above Hot Shutdown, was assigned an ECL of Site Area Emergency.
Precursor events to a loss of all AC power were also included as an Unusual Event and an Alert. A station blackout coping analyses performed in response to IO CFR 50.63 and Regulatory Guide 1.155, Station Blackout, may be used to determine a tirrie,..based criterion to demarcate between a Site Area Emergency and a General Emergency.
The time dimension is critical to a properly anticipatory emergency declaration since the goal is to maximize the time available for State and local officials to develop and implement offsite protective actions. 2. For severe core damage events, uncertainties exist in phenomena important to accident progressions leading to containment failure. Because of these uncertainties, predicting the status of containment integrity may be difficult under severe accident conditions.
This is why maintaining containment integrity alone following sequences leading to severe core damage is an insufficient basis for not escalating to a General Emergency.
: 3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containment bypass, a large Loss of Coolant Accident (LOCA) with early containment failure, a Station Blackout lasting longer than the PBNP coping period, and a reactor coolant pump seal failure. The generic EAL methodology needs to be sufficiently rigorous to address these sequences in a timely fashion. 10 
 
===3.2 TYPES===
OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS The NEI 99-01 methodology makes use of symptom-based, barrier-based and based ICs and EALs. Each type is discussed below. Symptom-based ICs and EALs are parameters or conditions that _are measurable over some range using plant instrumentation ( e.g., core temperature, reactor coolant level, radiological effluent, etc.). When one or more of these parameters or conditions are normal, reactor operators will implement procedures to identify the probable cause(s) and take corrective action. Fission product barrier-based ICs and EALs are the subset of symptom-based EALs that refer specifically to the level of challenge to the principal barriers against the release of radioactive material-from the reactor core to the environment.
These barriers are the fuel cladding, the reactor coolant system pressure boundary, and the containment.
The barrier-based ICs and EALs consider the level of challenge to each individual barrier -potentially lost and lost -and the total number of barriers under challenge.
Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance.
These include the failure of an automatic reactor scram/trip to shut down the reactor, natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release. 11 3.3 PBNP-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION The scheme's generic information is organized by Recognition Category in the following order.
* R -Abnormal Radiation Levels / Radiological Effluent -Section 6
* C -Cold Shutdown/
Refueling System Malfunction
-Section 7
* E :.. Independent Spent Fuel Storage Installation (ISFSI) -Section 8
* F -Fission Product Barrier -Section 9 *
* H -Hazards and Other Conditions Affecting Plant Safety -Section 10
* S -System Malfunction
-Section 11 Each Recognition Category section contains a matrix showing the ICs and their associated emergency classification levels. The following information and guidance is provided for each IC:
* ECL -the assigned emergency classification level for the IC.
* Initiating Condition
-provides a summary description of the emergency event or condition.
* Operating Mode Applicability
-Lists the modes during which the IC and associated EAL(s) are applicable (i.e., are to be_ used to classify events or conditions).
* Emergency Action Level(s)-Provides examples ofreports and indications that are considered to meet the intent of the IC. For Recognition Category F, the fission product barrier thresholds are presented in tables and arranged by fission product barrier and the degree <:>fbarrier challenge (i.e., potential loss or loss). This presentation method shows the synergism among the thresholds, and supports accurate assessments.
* Basis -Provides background information that explains the intent and application of the IC and EALs. In some cases, the basis also includes relevant source information and references.
12 3.4 IC AND EAL MODE APPLICABILITY The PBNP emergency classification scheme was developed recognizing that the applicability ofICs and EALs will vary with plant mode. For example, some based ICs and EALs can be assessed only during the power operations, startup, or hot standby/shutdown modes of operation when all fission product barriers are in place, and plant instrumentation and SAFETY SYSTEMS are fully operational.
In the cold shutdown and refueling modes, different symptom-based ICs and EALs will come into play to reflect the opening of systems for routine maintenance, the unavailability of some SAFETY SYSTEM components and the use of alternate instrumentation.
The following table shows which Recognition Categories are applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable iri the indicated modes. MODE APPLICABILITY MATRIX Recognition Category Mode R C E F H s Power Operations X X X X X Startup X X X X X Hot Standby X X X X X Hot Shutdown . X X X X X Cold Shutdown X X X X Refueling X X X X Defueled X X X X PBNP OPERATING MODES MODE TITLE REACTIVITY
%RATED AVERAGE CONDITION THERMAL REACTOR (Keff) POWERCa) COOLANT TEMPERATURE (OF) 1 Power Operation 20.99 >5 NA 2 Startup 20.99 :S 5 NA-3' Hot Standby < 0.99 NA 2350 4 Hot ShutdownlbJ
< 0.99 NA 350 > Tavg > 200 5 Cold Shutdown lDJ < 0.99 NA :S 200 6 Refueling lCJ NA NA NA NIA Defueled All fuel removed from the reactor vessel (full core offload during refueling or extended outage) (a) Excluding decay heat (b) All reactor vessel head closure bolts fully tensioned. ( c) One or more reactor vessel head closure bolts less than fully tensioned.
13 4 PBNP SCHEME DEVELOPMENT
 
===4.1 GENERAL===
DEVELOPMENT PROCESS The PBNP ICs and EALs were developed to be unambiguous and readily assessable.
The IC is the fundamental event or condition requiring a declaration.
The EAL(s) is the pre-determined threshold that defines when the IC is met. Useful acronyms and abbreviations associated with the PBNP emergency classification scheme are presented in Appendix A, Acronyms and Abbreviations.
Many words or terms used in the PBNP emergency classification scheme have specific definitions.
These words and terms are identified by being set in all capital letters (i.e., ALL CAPS). The definitions are presented in Appendix B, Definitions.
 
===4.2 CRITICAL===
CHARACTERISTICS When crafting the scheme, PBNP ensured that certain critical characteristics have been met. These critical characteristics are listed below. *
* The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information are consistent with industry guidance; while the actual wording may be different, the classification intent is maintained.
With respect to Recognition Category F, PBNP includes a user-aid to facilitate timely and accurate classification of fission product barrier losses and/or potential losses. The user-aid logic is consistent with the classification logic presented in Section 9.
* The ICs, EALs, Operating Mode Applicability criteria, Notes and Basis information are technically complete and accurate (i.e., they contain the information necessary to make a correct classification).
* EAL statements use objective criteria and observable values.
* ICs, EALs, Operating Mode Applicability and Note statements and formatting consider human factors and are user-friendly.
* The scheme facilitates upgrading and downgrading of the emergency classification where necessary.
* The scheme fadlitates classification of multiple concurrent events or conditions.
14 
 
===4.3 INSTRUMENTATION===
 
USED FOR-EALS PBNP incorporated instrumentation that is reliable and routinely maintained in accordance with site programs and procedures.
Alarms referenced in EAL statements are those that are the_most operationally significant for the described event or condition.
EAL setpoints are within the calibrated range of the referenced instrumentation, and consider any automatic instrumentation functions that may impact accurate EAL assessment.
In addition, EAL setpoint values do not use terms such as "off-scale low" or "off-scale high" since that type of reading may not he readily differentiated from an instrument failure. 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA Some of the criteria/values used in several EALs and fission product barrier thresholds may be drawn from PBNP's AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments.
Appropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to determine if an evaluation pursuant to 10 CFR 50.54( q) is required.
15 5 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS
 
===5.1 GENERAL===
CONSIDERATIONS When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information.
In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL. NRC regulations require the licensee to establish and maintain the capability to assess, -classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on
* implementing this requirement in NSIR/DPR-ISG-01, Interim Staff Guidance, Emergency Planning for Nuclear Power Plants. All emergency classification assessments should be based upon valid indications, reports or conditions.
A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy.
For example, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.
The validation of indications should be completed in a manner that supports timely emergency declaration.
For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
For EAL thresholds that specify a duration of the off-normal condition, the NRC expects that the emergency declaration process run concurrently with the specified threshold duration.
Once the off-normal condition has existed for the duration specified in the EAL, no further effort on this declaration is necessary-the EAL has been exceeded.
Consider as an example, the EAL "fire which is not extinguished
~ithin 15 minutes of detection." On receipt of a fire alarm, the plant fire brigade is dispatched to the scene to begin fire suppression efforts.
* If the fire brigade reports that the fire can be extinguished before the specified duration, the emergency declaration is placed on hold while firefighting activities continue.
If the fire brigade is successful in extinguishing the fire within the specified duration from detection, no emergency declaration is warranted based on that EAL. 16
* If the fire is still burning after the specified duration has elapsed, the EAL is exceeded, no further assessment is necessary, and the emergency declaration would be made promptly.
As used here, "promptly" means at the first available opportunity (e.g., if the Shift Manager is receiving an update from the fire brigade at the 15-minute mark, it is expected that the declaration will occur as the next action after the call e.Q.,ds ).
* If, for example, the fire brigade notifies the shift supervision 5 minutes after detection that the brigade itself cannot extinguish the fire such that the EAL will be met imminently and cannot be avoided, the NRC would not consider it a violation of the licensee's emergency plan to declare the event before the EAL is met (e.g., the, 15-minute duration has elapsed).
While a prompt declaration would be beneficial to public health and safety and is encouraged, it is.not required by regulation.
* In all of the above, the fire duration is measured from the time the alarm, indication, or report was first received by the plant operators.
Validation or confirmation . establishes that the fire started as early as the time of the alarm, indication, or report. A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that 1) the activity proceeds as planned and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component.
In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected.
Events or conditions of this type may be subject to the reporting requirements of 10 CFR 50.72. . The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded ( e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis.
In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available).
The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time ( e.g., maintain the necessary expertise on-shift).
While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary.
This scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition.
A similar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 17 
 
===5.2 CLASSIFICATION===
 
METHODOLOGY To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded.
The evaluation of an EAL( s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with plant procedures.
When assessing an EAL that specifies a time dur;:1.tion for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01.
 
===5.3 CLASSIFICATION===
 
OF MULTIPLE EVENTS AND CONDITIONS When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared.
For example:
* If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two different units, a Site Area Emergency should be declared.
There is no "additive" effect from multiple EALs meeting the same ECL. For example:
* If two Alert EALs are met, whether at one unit or at two different units, an Alert should be declared.
Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events. 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable.
If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not ~hen it was declared).
Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.
For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response.
In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 18 5.5 5.6 J ' CLASSIFICATION OF IMMINENT CONDITIONS Although EALs pro events or conditions vide specific thresholds, the Emergency Director must remain alert to that could lead to meeting or exceeding an EAL within a relatively (i.e., a change in the ECL is IMMINENT).
If; in the judgment of the meeting an EAL is IMMINENT, the emergency classification short period of time Emergency Director, should be made as if the EAL has been met. While applicable to all emergency classification levels, classification levels this approach is particularly important at the higher emergency since it provides additional time for implementation of protective_
measures.
EMERGENCY CLASS IFICATION LEVEL UPGRADING AND DOWNGRADING An _ECL may be do wngraded when the event or condition that meets the highest IC and s, and other site-specific downgrading requirements are met. If EAL no longer exist downgrading the EC L is deemed appropriate, the new ECL would then be based on a lower applicable IC( s) and EAL(s). The ECL may also simply be terminated.
The following appro ach to downgrading or terminating an ECL is recomme_nded.
ECL Unusual Event Alert Site Area Emergen cywithno mage long-term plant da Site Area Emergen long-term plant da cywith -mage General Emergency Action When Condition No Longer Exists Terminate the emergency in accordance with plant procedures.
Downgrade or terminate the emergency in accordance with plant procedures.
Downgrade or terminate the emergency in accordance with plant procedures.
Terminate the emergency and enter recovery in accordance with plant procedures.
Terminate the emergency and enter recovery in accordance with plant procedures.
As noted above, gui dance concerning classification of rapidly escalating events or d in RIS 2007-02. conditions is provide 19 ' 
 
===5.7 CLASSIFICATION===
 
OF SHORT-LIVED EVENTS As discussed in Section 3 .2, event-based I Cs and EALs define a variety of specific occurrences that have potential or actual safety significance.
By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed.
If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration.
Examples of such events include a failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip or an earthquake.
5.8
* CLASSIFICATION OF TRAN~IENT CONDITIONS Many of the I Cs and/or EALs contained in this document employ time-based criteria.
These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted.
In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes).
The following guidance should be applied to the classification of these conditions.
EAL momentarily met during expected plant response -In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems a;nd components are operating as expected, and operator actions are performed in accordance with procedures.
EAL momentarily met but the condition is corrected prior to an emergency declaration
-If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required.
For illustrative purposes, consider the following example. An A TWS occurs and the auxiliary feed water system fails to automatically start. Steam generator levels rapidly decrease and the plant enters an inadequate RCS . heat removal condition (a potential loss of both the fuel clad and RCS barriers).
If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal condition prior to an emergency declaration, then the classification should be based on the A TWS only. It is important to stress that the 15-minute emergency classification assessment period is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event; emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations where an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration.
This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
20 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition.
This situation can occur when persollilel discover that an event or condition existed which met an EAL, but no emergency-was declared, and the event or condition no longer exists at the time of discovery.
This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 is applicable.
Specifically, the event should be reported to the NRC in accordance with 10 CFR 50.72 within one hour of the discovery of the undeclared event or condition.
The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.
5 .10 RETRACTION OF AN EMERGENCY DECLARATION Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022
.. 21 rl 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS 22 RU1 ECL: Unusual Event Initiating Condition:
Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer. Operating Mode Applicability:
All Emergency Action Levels: Notes:
* The Emergency Director should declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
RUl.1 RUl.2 RUl.3 Reading on ANY of the following effluent radiation monitors greater than the reading shown for 60 minutes or longer: Monitor Reading 1(2)RE-307 CTMNT Purge Exhaust Mid-Range Gas 1.4E-2 &#xb5;Ci/cc with only containment purge in operation 2RE-305 CTMNT Purge Exhaust Low Range Gas 9.4E-3 &#xb5;Ci/cc with both purge and GS building ventilation in operation 2RE-307 CTMNT Purge Exhaust Mid-Range Gas 9.4E-3 &#xb5;Ci/cc with both purge and GS building ventilation in operation 2RE-307 CTMNT Purge Exhaust Mid-Range Gas 2.8E-2 &#xb5;Ci/cc with only GS building ventilation in operation 2RE-307 CTMNT Purge Exhaust Mid-Range Gas 1.0E+ 1 &#xb5;Ci/cc with only forced vent of containment 2RE-309 CTMNT Purge Exhaust High Range Gas l .OE+ 1 &#xb5;<;i/cc with only forced vent of containment RE-315 AB Exhaust Low Range Gas 5.4E-3 &#xb5;Ci/cc RE-317 AB Exhaust Mid-Range Gas 5.4E-3 &#xb5;Ci/cc RE-325 Drumming Area Exhaust Low Range Gas 8.4E-3 &#xb5;Ci/cc RE-327 Drumming Area Exhaust Mid-Range Gas 8.4E-3 &#xb5;Ci/cc 1(2)RE-229 Service Water Overboard 2.3E-3 &#xb5;Ci/cc -Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer. Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer. 23 Definitions:
None Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a level radiological release that exceeds regulatory commitments for an extended period of time ( e.g., an uncontrolled release).
It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.
PBNP incorporates design features intended to control the release of radioactive effluents to the environment.
Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases.
The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological' effluent EALs more fully addresses the spectrum of possible accident events and conditions.
Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Releases should not be prorated or averaged.
For example, a release exceeding 4 times release limits for 3 0 minutes does not meet the EAL. EAL RUl .1 -This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.
EAL RUl .2 -This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas). EAL RUl .3 -This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.). Escalation of the emergency classification level would be via IC RAl. 24 ECL: Unusual Event Initiating Condition:
UNPLANNED loss of water level above irradiated fuel. Operating Mode Applicability:
All Emergency Action Levels: RU2 RU2.l a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
* Spent fuel pool low water level alarm
* Visual observation AND b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
* RE-105 SFP Area Low Range Radiation Monitor
* RE-135 SFP Area High Range Radiation Monitor
* 1(2)RE-102 EL 66' CONTAINMENT Low Range Monitor Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient; The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal. Basis: This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. A water level decrease will be primarily determined by indications from available level instrumentation.
The low level alarm is actuated by LC-634, SFP Level Indicator at 62 ft 8 in. based OJ;l maintaining at least 6 ft. of water on a withdrawn fuel assembly.
Other sources oflevel indications may include reports from plant personnel ( e.g., from a refueling crew) or video camera observations (if available).
A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.
25 The effects of planned evolutions should be considered.
For example,-a refueling bridge area radiation monitor reading may increase due to planned evolutions suc:h as lifting of the reactor vessel head or movement of a fuel as_sembly.
Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level. A drop in water level above irradiated fuel within the reactor ve,ssel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC RA2. 26 RA1 ECL: Alert Initiating Condition:
Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Operating Mode Applicability:
All Emergency Action Levels: Notes:
* The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
* The pre-calculated effluent monitor values presented in EAL RAl.1 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
RAl.1 Reading on ANY of the following radiation monitors greater. than the reading shown for 15 minutes or longer: Monitor Reading 1(2)RE-307 CTMNT Purge Exhaust Mid-Range Gas 6.0E+O &#xb5;Ci/cc with only containment purge in operation 1(2)RE-309 CTMNT Purge Exhaust High Range Gas 6.0E+O &#xb5;Ci/cc with only containment purge in operation 2RE-307 CTMNT Purge Exhaust Mid-Range Gas 4.0E+O &#xb5;Ci/cc with both purge and GS building ventilation in operation 2RE-309 CTMNT Purge Exhaust High Range Gas 4.0E+O &#xb5;Ci/cc with both purge and GS building ventilation in operation 2RE-309 CTMNT Purge Exhaust High Range Gas 1.2E+ 1 &#xb5;Ci/cc with only GS building ventilation in operation RE-317 AB Exhaust Mid-Range Gas _1.0E+O &#xb5;Ci/cc RE-319 AB Exhaust High Range Gas 1.0E+O &#xb5;Ci/cc RE-327 Drumming Area Exhaust Mid-Range Gas 1.6E+O &#xb5;Ci/cc 27 ,
RAl.2 RAl.3 RAl.4 Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond SITE BOUNDARY.
Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for one hour of exposure.
Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.
Definitions:
SITE BOUNDARY:
That lii;ie beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.
Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 % of the EPA Protective Action Guides (PA Gs). It includes both monitored and un-monitored releases.
Releases of this magnitude represent an actual or potential substantial degradation of the level of-safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
This IC is modified by a note that EAL RAl.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 1 % of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
For EAL RAl.3, there are no site-specific liquid radiation monitors capable of monitoring 1iquid effluent releases at the classification threshold for this EAL because their detector operating range is exceeded prior to reaching these levels. Entry into this EAL for a liquid radioactivity release will be based on sampling initiated due to entry into EAL RUl. In practical terms, this means that entry into IC RUl will start sampling (per RMS Alarm Setpoint and Response Book) which will then allow detection of the setpoint for RAl. Escalation of the emergency classification level would be via IC RS 1. 28 RA2 ECL: Alert Initiating Condition:
Significant lowering of water level above, or damage to, irradiated fuel. Operating
_Mode Applicability:
All Emergency Action Levels: Uncovery of irradiated fuel in the REFUELING PATHWAY. RA2.l RA2.2 Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by a reading on ANY of the following radiation monitors greater than the value shown: Monitor Reading RE-105 SFP Area Low Range Radiation Monitor 4R/hr .... 1(2)RE-126 Containment High Radiation Monitor 7R/hr f-1(2)RE-127 Containment High Radiation Monitor 7R/hr f-1(2)RE-128 Containment High Radiation Monitor 7R/hr RA2.3 Lowering of spent fuel pool level to 49 ft.O in. Definitions:
REFUELING PATHWAY -The reactor refueling cavity, spent fuel pool and fuel transfer canal. Basis: This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment.
As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC EUl. Escalation of the emergency would be based on either Recognition Category R or C ICs.
* EALRA2.l This EAL escalates from RU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation ( e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters.
Computational aids may also be used. Classification of an event using this EAL should be based on the totality of available indications, reports, and observations.
29 
( While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in -some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered.
To the degree possible, readings should be considered in combination with other available indications of inventory loss. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. EALRA2.2 This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly.
A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event ( e.g., a fuel handling accident).
EALRA2.3 Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via I Cs RS 1 or RS2. 30 RA3 ECL: Alert Initiating Condition:
Radiation levels-that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability:
All Emergency Action Levels: Note: If the equipment in the listed room or area was already inoperable or out-of-service before the eve1;it occurred, then no emergency classification is warranted.
RA3.1 RA3.2 Dose rate greater than 15 mR/hr in ANY of the following areas:
* Control Room (RE-101)
* Central Ala~ Station AND Secondary Alarm Station (by survey) An UNPLANNED event results jn radiation levels that prohibit or impede access to any of the following plant rooms or areas: Area Mode Ul VCT Area 3/4/5 U2 VCT Area 3 /4 I 5 Ul Primary Sample area 3 U2 Primary Sample area 3 CCWHXRoom 4/5 C-59 area 3/4/5 Pipeway 2, 8 ft. Elev. 3/4 Pipeway 3, 8 ft. Elev. 3/4 l/2B32 MCC Area 4 Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown.
As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiatlon levels and determine if another IC may be applicable.
31 For EAL RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits). An emergency declaration is not warranted if any of the following conditions apply.
* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and-the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.
* The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area ( e.g., radiography, spent filter or resin transfer, etc.). *
* The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
* The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action. The list of plant rooms or areas in EAL RA3.2 was generated from a step-by-step review of OP-3A, 3B, 3C, SD, and 7 A. Escalation of the emergency classification level would be via Recognition Category R, C or F ICs. 32*
RS1 ECL: Site Area Emergency Initiating Condition:
Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE. Operating Mode Applicability:
All Emergency Action Levels: Notes:
* The Emergency Director should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
* The pre-calculated effluent monitor values presented in EAL RS 1.1 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
RSl.l Reading on ANY of the following radiation monitor greater than the reading shown for 15 minutes or longer: Monitor Reading 1(2)RE-307 CTMNT Purge Exhaust Mid-Range Gas 6.0E+l &#xb5;Ci/cc with only containment purge in operation 1(2)RE-309 CTMNT Purge Exhaust High Range Gas 6.0E+ 1 &#xb5;Ci/cc with only containment purge in operation 2RE-307 CTMNT Purge Exhaust Mid-Range Gas 4.0E+ 1 &#xb5;Ci/cc with both purge and GS building ventilation in operation 2RE-309 CTMNT Purge Exhaust High Range Gas 4.0E+ 1 &#xb5;Ci/cc with both purge and GS building ventilation in operation 2RE-309 CTMNT Purge Exhaust High Range Gas l.2E+2 &#xb5;Ci/cc with only GS building ventilation in operation RE-317 AB Exhaust Mid-Range Gas 1.0E+ 1 &#xb5;Ci/cc RE-319 AB Exhaust High Range Gas l .OE+ I &#xb5;Ci/cc RE-327 Drumming Area Exhaust Mid-Range Gas 1.6E+ 1 &#xb5;Ci/cc 33 RSl.2 RSl.3 Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY.
Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
* Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.
Definitions:
SITE BOUNDARY:
That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the 'licensee.
Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PA Gs). It includes both monitored and rm-monitored releases.
Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. This IC is modified by a note that EAL RS 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
Escalation of the emergency classification.
level would be via IC RG 1. 34 ECL: Site Area Emergency Initiating Condition:
Spent fuel pool level at 40 ft. 8 in. Operating Mode Applicability:
All Emergency Action Levels: RS2.1 Lowering of spent fuel pool level to 40 ft. 8 in .. Definitions:
None Basis: RS2 This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.
Escalation of the emergency classification level would be via IC RGl or RG2. 35 RG1 ECL: General Emergency Initiating Condition:
Release of gaseous radi9activity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE. Operating Mode Applicability:
All Emergency Action Levels: Notes:
* The Emergency Director should declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
* The pre-calculated effluent monitor values presented in EAL RG 1.1 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
RGl.1 RGl.2 RGl.3 Reading on ANY of the following radiation monitor greater than the reading shown for 15 minutes or longer: Monitor Reading 1(2)RE-309 CTMNT Purge Exhaust High Range Gas 6.0E+2. &#xb5;Ci/cc with only containment purge in operation 2RE-309 CTMNT Purge Exhaust High Range Gas 4.0E+2 &#xb5;Ci/cc with both purge and GS building ventilation in operation 2RE-309 CTMNT Purge Exhaust High Range Gas l.2E+3 &#xb5;Ci/cc with only GS building ventilation in 'operation RE-317 AB Exhaust Mid-Range Gas l.OE+2 &#xb5;Ci/cc RE-319 AB Exhaust High Range Gas l.OE+2 &#xb5;Ci/cc Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY.
Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:
*, Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or ionger.
* Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrein for one hour of inhalation.
36 Definitions:
SITE BOUNDARY:
That line beyond which the land is neither owned, nor leased, nor otherwise
* controlled by the licensee.
Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PA Gs). It includes b_oth monitored and un-monitored releases.
Releases of this magnitude will require implementation of protective actions for the public. This IC is modified by a note that EAL RG 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions.
Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.
The TEDE dose is set at the EPA P AG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established.
If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
37 ECL: General Emergency Initiating Condition:
Spent fuel pool level cannot be restored to at least 40 ft.8 in. for 60 minutes or longer. Operating Mode Applicability:
All Emergency Action Levels: Note: The Emergency Director should declare the General Emergency promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
RG2 RG2.1 Spent fuel pool level cannot be restored to at least 40 ft. 8 in. for 60 minutes or longer. Definitions:
None Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.
It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.
38 7 COLD SHUTDOWN/
REFUELING SYSTEM MALFUNCTION ICS/EALS 39 CU1 ECL: Unusual Event Initiating Condition:
UNPLANNED loss ofreactor vessel/RCS inventory for 15 minutes or longer. Operating Mode Applicability:
5, 6 Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
CUI.I CUl.2 UNPLANNED loss of reactor coolant results in reactor vessel/RCS level less than a required lower limit for 15 minutes or longer. a. Reactor vessel/RCS level cannot be monitored.
AND b. UNPLANNED increase in Containment Sump A OR Waste Holdup Tank levels. Definitions:
UNPLANNED:
A parameter ch:inge or an event that is not 1) the result of an inteuded evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inability to restore and maintain water level to a required minimum level ( or the lower limit of a level band), or a loss of the ability to monitor reactor vessel/RCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. , Refueling evolutions that decrease RCS water inventory are carefully planned and controlled.
An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. EAL CUI .1 recognizes that the minimum required reactor vessel/RCS level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented.
This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.
The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. 40 EAL CUI .2 addresses a condition where all means to determine reactor vessel/RCS level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative ofleakage from the reactor vessel/RCS.
Co:r;.tinued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAI or CA3. 41 CU2 ECL: Unusual Event Initiating Condition:
Loss of all but one AC power source to emergency buses for 15 minutes or longer. Operating Mode Applicability:
5, 6, Defueled Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded .. CU2.1 a. AC power capability to 1(2)A-05 and 1(2)A-06 is reduced to a single power source for 15 minutes or longer. AND b. Any additional single power source failure will result in loss-of all AC power to SAFETY SYSTEMS. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: Note: with respect to this EAL, "Station Blackout is Unit 1(2) specific." This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in ff loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than.one, train of safety-related equipment.
* Normal Unit 1(2) offsite power sources include: o 345 KVAC 1(2)X-93 through the 13.8 KVAC system to the LVSAT 1(2)X-04 o 345 KVAC backfed through the 19 KVAC system to the UAT 1(2)X-02
* Normal Unit 1(2) onsite power sources consist of: o emergency diesel generators o gas turbine generator o unit main turbine generator o power supplied from the opposite unit When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential,_degradat1on of the level of safety of the plant. 42 Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. An "AC power source" is a source recognized in AOPs and EOPs (including Beyond Design Basis event procedures), and capable of supplying required power to an emergency bus. Some examples of this Initiating Condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
* A loss of all offsite power and loss of all emergency power sources ( e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. 43 ECL: Unusual Event Initiating Condition:
UNPLANNED increase in RCS temperature.
Operating Mode Applicability:
5, 6 Emergency Action Levels: CU3 Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
CU3.l CU3.2 UNPLANNED increase in RCS temperature to greater than 200&deg;F. Loss of ALL RCS temperature and reactor vessel/RCS level indication for 15 minutes or longer. Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
Basis: This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
EAL CU3 .1 involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications.
During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled.
A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.
44 EAL CU3 .2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removat During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation.
Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation to Alert would be via IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
45 ECL: Unusual Event Initiating Condition:
Loss of Vital DC power for 15 minutes or longer. Operating Mode Applicability:
5, 6 Emergency Action Levels: CU4 Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
CU4.1 Indicated voltage is less than 115 VDC on required Vital DC buses D-01, D-02, D-03, or D-04 for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant. As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment.
For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train Bis in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification.
The safety-related 125 VDC system consist of four main buses; D-01, D-02, D-03, and D-04. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CAI or CA3, or an IC in Recognition Category R. 46 ECL: Unusual Event Initiating Condition:
Loss of all onsite or offsite communications capabilities._
Operating Mode Applicability:
5, 6, Defueled Emergency Action Levels: CU5.l Loss of ALL of the following onsite communication methods:
* Plant Public Address System (Gai-Tronics)
* Commercial Phones
* PBXPhones
* ~ecurity Radio
* Portable Radios cus CU5.2 Loss of ALL of the following offsite response organization communications methods:
* Nuclear Accident Reporting System (NARS)
* Commercial Phones
* PBXPhones
* Satellite Phones
* Manitowoc County Sheriffs Department Radio CU5.3 Loss of ALL of the following NRC communications methods:
* FTS Phone System
* Commercial Phones
* PBXPhones
* Satellite Phones Definitions:
None Basis: This IC addresses a significant loss of on-site or offsite communications capabilities.
While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to offsite response organizations and the NRC.
* This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). EAL CU5.1 addresses a total loss of the communications methods used in support of routine plant operations.
EAL CU5.2 addresses a total loss of the communications methods used to notify all offsite -response organizations of an emergency declaration.
The offsite response organizations referred to here are the State of Wisconsin, Manitowoc County, and Kewaunee County. EAL CU5.3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
47 ECL: Alert Initiating Condition:
Loss ofreactor vessel/RCS inventory.
Operating Mode Applicability:
5, 6* Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
CAI.I CAl.2 Loss of reactor vessel/RCS inventory as indicated by level less than 16% on LI-447 / LI-447A. . a. Reactor vessel/RCS level cannot be monitored for 15 minutes or longer AND CA1 b. UNPLANNED increase in Containment Sump A OR Waste Holdup Tank levels due to a loss of reactor vesseVRCS inventory.
Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses conditions that are precm;sors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier).
This condition represents a potential substantial reduction in the level of plant safety. For EAL CAI.I, a lowering of water level below 16% on LI-447 / LI-447A indicates that operator actions have not been successful in restoring and maintaining reactor vessel/RCS water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.
The LI-447/LI-447A threshold corresponds to the minimum shutdown reactor vessel level required for operation of RHR without air binding the suction. Although related, EAL CAI.I is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal ( e.g., loss of a Residual Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. For EAL CAl .2, the inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation_.
If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative ofleakage from the reactor vessel/RCS.
48 The 15-minute duration for the loss oflevel indication was chosen because it is half of the EAL duration specified in IC CS 1 If the reactor vessel/RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS 1. 49 CA2 ECL: Alert Initiating Condition:
Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer. Operating Mode Applicability:
5, 6, Defueled Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
CA2.1 Loss of ALL offsite and ALL onsite AC Power to 1(2)A-05 and 1(2)A-06 for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant. For the purpose of classification under this EAL, evaluation of power sources should be made on each unit individually.
If mitigative strategies establish emergency power to any bus listed in the EAL, the EAL threshold for this Initiating Condition is not met. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS 1 or RS 1. 50 CA3 ECL: Alert Initiating Condition:
Inability to maintain the plant in cold shutdown.
Operating Mode Applicability:
5, 6 Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
CA3.l CA3.2 UNPLANNED increase in RCS temperature to greater than 200&deg;F for greater than the duration specified in the following table. Table: RCS Heat-up Duration Thresholds RCS Status CONTAINMENT CLOSURE Heat-up Duration Status Intact (but not at reduced Not applicable 60 minutes* inventory) " Not intact ( or at reduced Established 20 minutes* inventory)
Not Established 0 minutes
* IfRHR is in operation within this time franie and RCS temperature is being reduced, the EAL is not applicable.
UNPLANNED RCS pressure increase greater than 25 psig. (This EAL does not apply during water-solid plant conditions.)
Definitions:
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
Basis: This IC addresses conditions involving a losss.of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.
51 The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation in PWRs). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety. Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes).
This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel. EAL CA3.2 provides a pressure-based indication of RCS heat-up. Escalation of the emergency classification level would be via IC CSl or RSI. 52 CA6 ECL: Alert Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability:
5, 6 Emergency A_ction Levels: Notes: CA6.1
* If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
* If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
: a. b. The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external :flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Lake level greater than or equal to 9.0 ft. (Plant elevation)
* Other events with similar hazard characteristics_
as determined by the Shift Manager or Emergency Director AND 1. Event damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode. -AND 2. EITHER of ~he following:
* Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, or
* The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. 53 Definitions:
FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
EXPLOSION:
A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction, or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure ( caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to* determine if the attributes of an explosion are present. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the ha:z:ardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria CA6.1.b.1 of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY .SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC RS 1. 54 CS1 ECL: Site Area Emergency Initiating Condition:
Loss of reactor vessel/RCS inventory affecting core decay heat removal* capability.
Operating Mode Applicability:
5, 6 Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
CSI.1 a. Reactor vessel/RCS level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by ANY of the following:
Definitions:
* Containment High Radiation Monitor, 1(2)RE-I26, 1(2)RE-I27, or 1(2)RE-I28, reading greater than 100 R/hr .
* Erratic source range monitor indication
* UNPLANNED increase in Containment Sump A OR Waste Holdup Tank levels of sufficient magnitude to indicate core uncovery UNPLANNED:
A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses a significant and prolonged loss of reactor vessel/RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.
Following an extended loss of core decay heat .removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.
In EAL CS I.I.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
55 For EAL CSl.1.a, the calculated radiation level on the Containment High Radiation Monitors (RE-126, RE-127, or RE-128) is without the reactor head in place. Calculated radiation levels
* with the reactor head in place are below the usable scale of these monitors.
The inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, equipment not calibrated for the plant conditions (e.g., hot cal only), or water level dropping below the range of available instrumentation.
If water level cannot be monitored, operators may determine-that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS.
These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
Escalation of the emergency classification level would be via IC CG 1 or RG 1. 56 CG1 ECL: General Emergency Initiating Condition:
Loss ofreactor vessel/RCS inventory affecting fuel clad integrity with containment challenged.
Operating Mode Applicability:
5, 6 Emergency Action Levels: Note: The Emergency Director should declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
CGl.1 a. Reactor vessel/RCS level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by ANY of the following:
J
* Containment High Radiation Monitors, 1(2)RE-126, 1(2)RE-127, or 1(2)RE-128, reading greater than ioo R/hr
* Erratic source range monitor indication
* UNPLANNED increase in Containment Sump A OR Waste Holdup Tank levels of sufficient magnitude to indicate core uncovery AND c. ANY indication from the Containment Challenge Table C-1. Containment Challenge Table C-1
* CONTAINMENT CLOSURE not established*
* 6% H 2 exists inside containment
* UNPLANNED increase in containment pressure
* If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute time limit, then declaration of a General Emergency is not required.
57 ---
Definitions:
CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment and it$ associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged.
This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.
Releases can be reasonably expected to exceed EPA P AG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.
With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment.
If CONTAINMENT CLOSURE is established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen bum (i.e., at the lower deflagration limit}. A hydrogen bum will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.
It therefore represents a challenge to Containment integrity.
In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment.
If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.
In EAL CG 1.1.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties).
It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
For EAL CGI.l.b, the calculated.radiation level on the Containment High Radiation Monitors, 1(2)RE-126, 1(2)RE-127, or 1(2)RE-128, is without the reactor head in place. Calculated radiation levels with the reactor head in place are below the usable scale of these monitors.
58 The inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, equipment not calibrated for the plant conditions ( e.g., hot cal only), or water level dropping below the range of available instrumentation.
If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes* niust be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS.
These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.
59 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS 60 ISFSI MALFUNCTION ECL: Unusual Event Initiating Condition:
Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability:
All Emergency Action Levels: EU1 EUl.1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiatton reading greater than the values shown below on the surface of the spent fuel cask. 32PTDSC Front Surface 1700 mrem/hr Door Centerline 400 mrem/hr End Shield Wall Exterior 12 mrem/hr VSC-24 Sides 200 mrem/hr Top 400 mrem/hr Air Inlets 700 mrem/hr Air Outlets 200 mrem/hr Definition:
CONFINEMENT BOUNDARY:
The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. Basis:. This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category RIC RUl, is used here to distinguish between non-emergency and emergency conditions.
The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSis are covered under ICs HUl and HAl. 61 9 FISSION PRODUCT BARRIER ICS/EALS 62 
: 1. A. 2. Table 9-F-1: EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FAlALERT FS1 SITE AREA EMERGENCY FGlGENERALEMERGENCY Any Loss or any Potential Loss of either Loss or Potential Loss of any two barriers.
Loss of any two barriers and Loss or the Fuel Clad or RCS barrier. Potential Loss of the third barrier. Operating Mode Applicability:
1, 2, 3, 4 Operating Mode Applicability:
1, 2, 3, 4 Operating Mode Applicability:
1, 2, 3, 4 . . ** .. ** RCS Barrier * . . Containment Barrier FueLClad Barrier . 1,, *. LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS Critical Safety Function Status 1. Critical Safety Function Status 1. Critical Safety Function Status Conditions A. Conditions Not Applicable A. Conditions Not Appli~able A. 1. Conditions requiring entry requiring entry into requiring entry into requiring entry into Core Cooling Core Cooling Heat Sink RED into Core* RED Path (CSP ORANGE Path Path (CSP H.1) are Cooling RED C. l) are met. (CSP C.2) are met. met. Path (CSP C.1) OR OR are met. B. Conditions B. Conditions AND requiring entry into requiring entry into 2. CSP C.1 not Heat Sink RED ' RCS Integrity effective within Path (CSP H.1) are RED Path (CSP 15 minutes. met. P.1) are met. RCS or SG Tube Leakage 2. RCS or SG Tube Leakage 2. RCS or SG Tube Leakage Not Applicable Not Applicable A. An automatic or A. Operation of a A. A leaking or Not Applicable manual ECCS (SI) standby charging RUPTURED SG is actuation is required (makeup) pump is FAULTED outside of by EITHER of the required by containment.
following:
EITHER of the 1. UNISOLABLE following:
RCS leakage 1. UNISOLABLE OR RCS leakage 2. SG tube OR RUPTURE. 2. SG tube leakage. 63 
: 3. RCS Activity / Containment Radiation
: 3. RCS Activity/
Containment Radiation
: 3. RCS Activity/
Containment Radiation A Containment Not Applicable A. Containment Not Applicable Nof Applicable A. Containment radiation monitor radiation monitor radiation monitor ' reading greater reading greater than reading greater than than 577 R/hr 11 R/hr indicated on 18,500 R/hr indicated on ANY ANY of the indicated on ANY . of the following.
following:
of the following.
* 1(2)RE-126
* 1(2)RE-126
* 1(2)RE-126
* 1(2)RE-127
* 1(2)RE-127
* 1(2)RE-127
* 1(2)RE-128
* 1(2)RE-128
* 1(2)RE-128 OR B. 1(2)RE-109 greater than 4,500 mR/hr 4. Containment Integrity or Bypass 4. Containment Integrity or Bypass 4. Containment Integrity or Bypass Not AppHcable Not Applicable Not Applicable Not Applicable A. Containment isolation A. Containment is required pressure greater AND EITHER of the than 60 psig. following:
OR 1. Containment B. 6% H 2 inside integrity has been containment.
lost based on OR Emergency C. 1. Containment Director pressure greater judgment.
than 25 psig. OR AND 2. UNISOLABLE
: 2. Less than one pathway from the full train of containment to depressurization the environment exists. equipment is operating per OR design for B. Indications of RCS
* 15 minutes or leakage outside of longer. containment.
64 
: 5. Emergency Director Judgment 5. Emergency Director Judgment 5. Emergency Director Judgment . A. ANY condition in A. ANY condition in A. ANY condition in the A. ANY condition in A. ANY condition in the A. ANY condition in the opinion of the the opinion of the I opinion of the . the opinion of the opinion of the the opinion of the Emergency Emergency Emergency Director Emergency Emergency Director Emergency Director Director that Director that that indicates Loss of Director that that indicates Loss of that indicates indicates Loss of indicates l'otential the RCS Barrier. indicates Potential the Containment Potential Loss of the Fuel Clad Loss of the Fuel Loss of the RCS Barrier.
* the Containment Barrier. Clad Barrier. Barrier. Barrier. 65 Basis Information For Fission Product Barrier Table 9-F-1 FUEL CLAD BARRIER THRESHOLDS:
The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. 1. Critical Safety Function Status Loss I.A This reading indicates temperatures within the core are sufficient to cause significant superheating ofreactor coolant. Core Cooling -RED indicates significant superheating and core uncovery and is considered to indicate loss of the Fuel Clad Barrier. CSP-C. l is the Critical Safety Procedure that provides directions to restore core cooling. Potential Loss 1.A This reading indicates temperatures within the core are sufficient to allow the onset of heat-induced cladding damage. Core Cooling -ORANGE indicates subcooling has been lost and that some clad damage may occur. CSP-C.2 is the Critical Safety Procedure that provides directions to restore adequate core cooling. Potential Loss 1.B This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. Verification of the EALs for CSP-H.1 should include an assessment of if feed flow was reduced based on the actions performed for an uncontrolled depressurization of both steam generators.
If this is the case, then declaration requirements should not be considered to be met. This does not affect the time that the CSFST initially changed color and met the CSP entry conditions for potential EAL event classification.
Alternatively, if CSP-H.1 was entered during a loss of coolant accident, it may be exited if secondary heat sink is not required based on RCS pressure less than non-faulted S/G pressure.
If this is the case, then declaration requirements should not be considered to be met. Meeting this threshold results in a Site Area Emergency because this threshold is identical to RCS Barrier Potential Loss threshold 2.A; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted.
66 FUEL CLAD BARRIER THRESHOLDS:
: 2. RCS or SG Tube Leakage There are no Loss or Potential Loss thresholds associated with RCS or SG Tube Leakage. 3. RCS Activity / Containment Radiation Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300&#xb5;Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 3.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.
Loss 3.B This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. There is no Potential Loss threshold associated with RCS Activity/
Containment Radiation.
: 4. Containment Integrity or Bypass Not Applicable (included for numbering consistency)
: 5. Emergency Director Judgment Loss 5.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost. Potential Loss 5.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
67 RCS BARRIER THRESHOLDS:
The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves. 1. Critical Safety Function Status There is no Loss threshold associated with Critical Safety Function Status. Potential Loss 1.A This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. If CSP-H.l was entered during a loss of coolant accident, it may be exited if secondary heat sink is not required based on RCS pressure less than non-faulted S/G pressure.
If this is the case, then declaration requirements should not be considered to be met. Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold l .B; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system. CSP-H.l is the Critical Safety Procedure that provides directions if the ultimate heat sink function is under extreme challenge.
This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature and thus a Potential Loss of the RCS barrier. In accordance with EOPs, there may be unusual accident conditions during which operators in!entionally reduce the heat removal capability of the steam generators; during these conditions, classification using this threshold is not warranted.
Potential Loss 1.B This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock-a transient that.causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).
CSP-P.1 is the Critical Safety Procedure that provides directions to avoid, or limit, thermal shock or pressurized thermal shock to the reactor pressure vessel or overpressurization conditions at low temperatures.
IF CSP-P.1 is entered duiing a large break loss of coolant accident, it may be exited if RCS pressure is low enough to allow for RHR forward flow. If CSP-P.1 is exited for this reason, then the Potential Loss criteria should not be considered met. 68 RCS BARRIER THRESHOLDS:
: 2. RCS or SG Tube Leakage Loss 2.A This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier. This threshold is applicable to-unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location -inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED.
If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold I .A will also be met. Potential Loss 2.A This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI) actuation has not occurred.
The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer level. This threshold is applicable to unidentified and pressure boundary leakage, as well as* identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location -inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
If a leaking steam generator is also FAUL TED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold I .A will also be met. 3. RCS Activity / Containment Radiation Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 3.A since it indicates a loss of the RCS Barrier only. There is no Potential Loss threshold associated with RCS Activity/
Containment Radiation.
69 RCS BARRIER THRESJIOLDS:
: 4. Containment Integrity or Bypass Not Applicable (included for numbering consistency)
: 5. Emergency Director Judgment Loss 5.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost. Potential Loss 5.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
70 CONTAINMENT BARRIER THRESHOLDS:
The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
: 1. Critical Safety Function Status Tree There is no Loss threshold associated with CSFST. Potential Loss I .A This condition repres 0 ents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure( s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier. The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing.
Whether or not the procedure(s) will be. effective should be apparent within 15 minutes. The Emergency Director should escalate the emergency classification level as soon as it is determined that the procedure( s) will not be effective.
Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.
: 2. RCS or SG Tube Leakage Loss 2.A This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment.
The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss I .A and Loss I .A, respectively.
This condition represents a bypass of the containment barrier. FAULTED is a defined term within the NEI 99-01 methodology; this determination is not necessarily dependent upon entry into, or diagnostic steps within,* an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably
[part of the FAULTED definition]
and the faulted steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAUL TED for emergency classification purposes.
71 CONTAINMENT BARRIER THRESHOLDS:
The FA UL TED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification.
Steam releases of this size are readily observable with normal Control Room indications.
The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuel clad barrier (i.e., RCS activity values) and IC SU5 for the RCS barrier (i.e., RCS leak rate values). This threshold also applies to prolonged steam releases necessitated by operational considerations such as t4e forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steani to the environment (and are thus similar to a FAULTED condition).
The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.
Steam releases associated with the expected operation of a SG atmospheric dump valve or safety relief valve do not meet the intent of this threshold.
Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown.
Steam releases associated with the unexpected operation of a valve ( e.g., a stuck-open safety valve) do meet this threshold.
Fallowing an SG tube leak or rupture, there may be minor radiological releases.
through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs. The emergency classification levels resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below. Affected SG is FAULTED Outside of Containment?
P-to-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU4 Unusual Event per SU4 Requires operation of a standby charging (makeup) Site Area Emergency Alert per F Al pump (RCS Barrier Potential per FSl Loss) Requires an automatic or Site Area Emergency manual ECCS (SI) actuation Alert per FAl (RCS Barrier Loss) per FSl There is no Potential Loss threshold associated with RCS or SG Tube Leakage. 72 CONTAINMENT BARRIER THRESHOLDS:
: 3. RCS Activity / Containment Radiation There is no Loss threshold associated with RCS Activity/
Containment Radiation.
Potential Loss 3 .A ..* The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into_ the containment, assuming that 20% of the fuel cladding has failed. Thislevel of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.
NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release ofradioactivity requiring offsite protective actions. For this condition to exist there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.
: 4. Containment Integrity or Bypass Loss 4.A These thresholds address a situation where containment isolation is required and one of two condition~
exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both thresholds 4.A.1 and 4.A.2. 4.A. l -Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage ( or sometimes referred to as design leakage (La)} Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may ( or may not) be accompanied by a noticeable drop in containment pressure.
Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency Director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors*
outside containment, operating status of containment pressure control equipment, etc.).
* Refer to the middle piping run of Figure 9-F-2. Two simplified examples are provided.
One is leakage froni a penetration and the other is leakage from an in-service system valve. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure. 73 CONTAINMENT BARRIER THRESHOLDS:
Another example would be a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment ( e.g., on a steam or feedwater line) and the other outside of containment.
In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable ( design) containment leakage through various penetrations or system components.
These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs. 4.A.2 -' Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment.
As used here, the term "environment" includes the atmosphere of a mom or area, outside the containment, that may, in turn, communicate with the outside-the-plant a.tmosphere ( e.g., through discharge of a ventilation system or atmospheric leakage).
Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
Refer to the top piping run of Figure 9-F-2. In this simplified example, the inboard and outboard isolation valves remained open after a containment isolation was required (i.e., containment isolation was not successful).
There is now an UNISOLABLE pathway from the containment to the environment.
The existence of a filter is not considered in the threshold assessment.
Filters do not remove fission product noble gases. In addition, a filter could b~come ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
Refer to the bottom piping run of Figure 9-F-2. In this simplified example, leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building.
The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump or system piping developed a leak that allowed steam/water to enter the Auxiliary Building, then threshold 4.B would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A.1 to be met as well. '-74 CONTAINMENT BARRIER THRESHOLDS:
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable ( design) containment leakage through various penetrations or system components.
Minor releases may also occur if a containment isolation valve( s) fails to close but the containment atmosphere escapes to a closed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs. The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold 2.A. Loss 4.B Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment.
If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence).
Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment should be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment.
If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly; however, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment.
Refer to the middle piping run of Figure 9-F-2. In this simplified example, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A.1 to be met as well. To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold 2.A to be met. Potential Loss 4.A If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. Thus, this threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier. 75 CONTAINMENT BARRIER THRESHOLDS:
Potential Loss 4.B
* The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity.
It therefore represents a potential loss of the Containment Barrier. Potential Loss 4.C This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may_ not have automatically started, if possible.
This threshold represents a potential loss of containment in that containment heat removal/depressurization systems (e.g., containment sprays, containment accident fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner. During a design basis accident, a minimum of two CONTAINMENT Accident Fan Cooler Units with their accident fans running and one CONTAINMENT spray train are required to maintain the CONTAINMENT peak pressure and temperature below the design limits. Each CONTAINMENT Spray train is a CONTAINMENT spray pump, spray header, nozzles, valves and piping. Each CONTAINMENT Accident Fan Cooler Unit consists of cooling coils, accident backdraft damper, accident fan, service water outlet valves, and controls necessary to ensure an operable service water flow path. 5. Emergency Director Judgment Loss 5.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost. Potential Loss 5.A ; This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
76 Inside Containment Damper RCP Seal Cooling Figure 9-F-2: Containment Integrity or Bypass Examples 1----------1 : Effluent : Auxiliary Building I I I Monitora ~---------
I I : Process : : Mon_itor : 77 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 78 ECL: Unusual Event Initiating Co~dition:
Confirmed SECURITY CONDITION or threat. Operating Mode Applicability:
All Emergency Action Levels: HU1 HUl.1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervisor.
HUl.2 HUl.3 Notification of a credible security threat directed at PBNP. A validated notification from the NRC providing information of an aircraft threat. Definitions:
SECURITY CONDITION:
Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION. HOSTILE ACTION: An act toward PBNP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent *may be included.
HOSTILE ACTION should not be construed to_include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may i.nclude violent acts between individuals in the owner controlled area). SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR 73.71 or 10 CFR 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAI, HSI and HGl. Timely and accurate communications between Security Shift Supervisor and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and offsite response organizations.
Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
79 EAL HUl .1 references Security Shift Supervisor because these are the individuals trained to confirm that a security event is occurring or has occurred.
Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR &sect; 2.390 information.
EAL HUl .2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with SY-AA-102-1014, Threat Assessment and Reporting.
EAL HUl .3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Offi_cer (HOO) will communicate to the licensee if the threat involves an aircraft.
The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with AOP-29, Security Threat. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HAl. 80 ECL: Unusual Event Initiating Condition:
Seismic event greater than OBE levels. Operating Mode Applicability:
All Emergency*
Action Levels: HU2 HU2.l Seismic event greater than Operating Basis Earthquake (OBE) as indicated by seismic monitor indication of ground acceleration greater than:
* 0.06 g horizontal OR
* 0.04 g vertical.
Definitions:
None Basis: This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE)1. An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE)2 should have no significant impact on related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant ( e.g., performs walk-downs and post-event inspections).
Given the time necessary to perform walk-downs and inspections, and fully understand .any impacts, this event represents a potential degradation of the level of safety of the plant. Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event. The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9. 1 An OBE is vibratory ground motion for which those features of a nuclear power plant necessary for continued operation without undue risk to the health and safety of the public will remain functional.
2 An SSE is vibratory ground motion for which certain (generally, safety-related) structures, systems, and components must be designed to remain functional.
81 HU3 ECL: Unusual Event Initiating Condition:
Hazardous events Operating Mode Applicability:
All Emergency Action Levels: Note: EAL HU3.4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
HU3.1 HU3.2 HU3.3 HU3.4 HU3.5 HU3.6 A tornado strike within the PROTECTED AREA. Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode. Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials ( e.g., an offsite chemical spill or toxic gas release).
A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.
Lake level greater than or equal to +8.0 ft. (Plant elevation).
Pump bay level less than -15.0 ft. Definitions:
PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL HU3.1 addresses a tornado striking (touching down) within the Protected Area. EAL HU3.2 addresses flooding of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns.
Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source ( e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. 82 EAL HU3 .3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA. EAL HU3.4 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles.
Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. EAL HU3.5 addresses lake water level as the Turbine Building is susceptible to external flooding.
+8.0 ft. corresponds to the Turbine Building floor elevation.
EAL HU3.6 addresses operability of the Emergency Diesel Generators and all Containment fan coolers. The pump bay level of -15.0 ft. represents the value at which the Emergency Diesel Generators and all Containment Fan Coolers must be declared inoperable, and is four feet above the level at which the service water pumps may begin to cavitate.
Escalation of the emergency classification level would be based on I Cs in Recognition Categories R, F, Sor C. 83 HU4 ECL: Unusual Event Initiating Condition:
FIRE potentially degrading the level of safety of the plant. Operating Mode Applicability:
All Emergency Action Levels: Notes:
* The Emergency Director should declare the Unusual Event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* A Containment fire alarm is considered valid upon receipt of multiple zones (more J than 1) on the FACP system (this note is applicable in Modes 1 and 2 only). HU4.1 a. A FIRE is NOT extinguished within IS-minutes of ANY of the following FIRE HU4.2 detection indications:
* Report from the field (i.e., visual observation)
* Receipt of multiple (more than 1) fire alarms or indications
* Field verification of a single fire alarm AND b. The FIRE is located within ANY Table H-1 plant rooms or areas a, Receipt of a single fire alarm with no other indications of a FIRE. AND b. The FIRE is located within ANY Table H-1 plant rooms or areas except Containment in Modes 1 and 2 (see Note above): AND c. The existence of a FIRE is not verified within 30-minutes of alarm receipt. HU4.3 A FIRE within the plant or ISFSI PROTECTED AREA not extinguished within 60 minutes of the initial report, alarm or indication.
HU4.4 A FIRE within the plant or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
Table H-1 Areas
* Control Room Containment PAB G05 building 13.8kV Building Cable Spreading Room Vital Switchgear Room AFW Pump Room G-01/02 Rooms EDG Building Service Water Pump Rooms Fa<;ade 85' 84 Definitions:
FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. Basis: This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. With regard to containment fire alarms, there is constant air movement in containment due to the operation of the air handling system drawing air to the cooling units past the smoke detectors.
It can reasonably be expected that a fire that burns for 15 minutes would produce sufficient products of combustion to cause fire detectors in multiple zones to alarm. EALHU4.1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished ( e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a
* subsequent verification action was performed.
Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. EALHU4.2 This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.
A single fire alarm, absent other indication( s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. Except for the Containment Building in Modes 1 or 2, the 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however,.
after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.
85 If an actual FIRE is verified by a report from the field, then EAL HU4. l is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.
EALHU4.3 In addition to a FIRE addressed by EAL HU4.1 or EAL HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of the ISFSI. EALHU4.4 If a FIRE within the plant or ISFSI PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency ( e.g., a local town Fire Department), then the level of plant safety is potentially degraded.
The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish.
Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Basis-Related Requirements from Appendix Rand NFPA-805 Criterion 3 of Appendix A to 10 CFR 50 states in part that "structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." The Nuclear Safety Goal ("NSG") in NFPA 805, Section 1.3.1 states, "The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance*because a safe shutdown success path, free of fire damage, must be available to nieet the nuclear safety goals, objectives and performance criteria for a fire under any plant operational mode or configuration.
*
* Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.
86 In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of I-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). Even though PBNP has adopted the alternate approach provided by NFPA-805 in lieu of the deterministic requirements of Appendix R, the 30-minutes to verify a single alarm as used in EAL HU4.2 is considered a reasonable amount of time to determine if an actual FIRE exists without presenting a challenge to the nuclear safety performance criteria.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9. 87 HU7 ECL: Unusual Event Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director . warrant declaration of a UE. Operating Mode Applicability:
All Emergency Action Levels: HU7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a UE. 88 HA1 ECL: Alert Initiating Condition:
HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes. Operating Mode Applicability:
All Emergency Action Levels: HAI.I HAl.2 A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor.
A validated notification from NRC of an aircraft attack threat within 30 minutes of the site. Definitions:
HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION: An act toward PBNP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). OWNER CONTROLLED AREA: The site property owned by or otherwise under the control* of the licensee.
PROJECTILE:
An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications between Security Shift Supervisor and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).
89 The Alert declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions. This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR 73.71 or 10 CFR 50.72. EAL HAl.1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA. EAL HAI .2 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and offsite response organizations are in a heightened state of readiness.
This EAL is met when the threat-related information has been validated in accordance with SY-AA-102-1014, Threat Assessment and Reporting.
The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.
The status and size of the plane may be provided by NORAD through the NRC. In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HS 1. 90 HAS ECL: Alert Initiating Condition:
Gaseorn:;release impeding access to equipment necessary for normal plant operations, cooldown, or shutdown.
Operating Mode Applicability:
All Emergency Action Levels: Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.
HAS.I a. Release of a toxic, corrosive, asphyxiant or flammable gas into any of the following plant rooms or areas: Area Mode Ul VCT Area 3/4/5 U2 VCT Area 3 /4 / 5 Ul Primary Sample area 3 U2 Primary Sample area 3 CCWHXRoom 4/5 C-59 area 3/4/5 Pipeway 2, 8 ft. Elev. 3/4 Pipeway 3, 8 ft. Elev. 3/4 1/2B32 MCC Area 4 AND b. Entry into the room or area is prohibited or impeded. Definitions:
None Basis: This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown.
This condition represents an actual or potential substantial degradation of the level of safety of the plant. An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release. 91 Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
An emergency declaration is not warranted if any of the following conditions apply.
* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release).
For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown.
do not require entry into the affected room until Mode 4.
* The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area ( e.g., fire suppression system testing).
* The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
* The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action. An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most c_ommonly, asphyxiants work by merely displacing air in an enclosed environment.
This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness, or even death. The list of plant rooms or areas in EAL HAS .1 was generated from a step-by-step review of OP-3A, 3B, 3C, 5D, and 7A. Escalation of the emergency classification level would be via Recognition Category R, C or F ICs. 92 HAG ECL: Alert Initiating Condition:
Control Room evacuation resulting in transfer of plant control to alternate locations.
Operating Mode Applicability:
All Emergency Action Levels: HA6.l An event has resulted in plant control being transferred from the Control Room to AOP local control stations.
Definitions:
None Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety. Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations.
The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel.
Activation of the ERO and emergency response facilities will assist in responding to these challenges.
Escalation of the emergency classification level would be via IC HS6. 93 HA7 ECL: Alert Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. Operating Mode Applicability:
All Emergency Action Levels: HA7.1 Other conditions exist which, in the judgment *of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Definitions:
HOSTILE ACTION: An act toward PBNP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert. 94 ECL: Site Area Emergency Initiating Condition:
HOSTILE ACTION within the PROTECTED AREA. Operating Mode Applicability:
All Emergency Action Levels: HS1 HSl.1 A HOSTILE ACTION is occurring or has occurred within the J;>ROTECTED AREA as reported by the Security Shift Supervisor.
Definitions:
HOSTILE ACTION: An act toward PBNP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to q.eliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.
INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. PROJECTILE:
An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. Basis: This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.
Timely and accurate communications between Security Shift Supervisor and the Control Room is essential for proper classification of a security-related event. 95 Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
As time and conditions allow, these events require a heightened state ofreadiness by the plant staff and implementation of onsite protective measures ( e.g., evacuation, dispersal or sheltering).
The Site Area Emergency declaration will mobilize offsite response organization resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.
This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HAI. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR 73.71 or 10 CFR 50.72. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.
This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat locatiol)..
Security-sensitive information should be contained in non-public documents such as the Security Plan. Escalation of the emergency classification level would be via IC HG 1. 96 r' .
HS6 ECL: Site Area Emergency Initiating Condition:
Inability to control a key safety function from outside the Control Room. Operating Mode Applicability:
All Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
HS6.1 a. An event has resulted in plant control being transferred from the Control Room to AOP local control stations.
AND b. Control of ANY of the following key safety functions is not reestablished within 15 minutes.
* Reactivity control
* Core cooling
* RCS heat removal Definitions:
None Basis: This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Director judgment.
The Emergency Director is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).
Escalation of the emergency classification level would be via IC FG 1 or CG 1. 97 HS7 ECL: Site Area Emergency Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.
Operating Mode Applicability:
All Emergency Action Levels: HS7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
Defmitions:
HOSTILE ACTION: An act toward PBNP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a Site Area Emergency.
98 HG1 ECL: General Emergency Initiating Condition:
HOSTILE ACTION resulting in loss of physical control of the facility.
Operating Mode Applicability:
All Emergency Action Levels: HGl.1 a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervisor.
AND b. EITHER of the following has occurred:
: 1. ANY of the following safety functions cannot be controlled or maintained.
* Reactivity control
* Core cooling
* RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT.
Definitions:
HOSTILE ACTION: An act toward PBNP or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped witli suitable weapons capable of killing, maiming, or . causing destruction.
* IMMINENT:
The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. 99 Basis: This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions.
It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system ( e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained.
Timely and accurate communications between Security Shift Supervisor and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.
Security-sensitive information should be contained in non-public documents such as the Security Plan. 100 HG7 ECL: General Emergency Initiating Condition:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.
Operating Mode Applicability:
All Emergency Action Levels: HG7.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or Ilv1MINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
* Definitions:
HOSTILE ACTION: An act toward PBNP or its personnel that includes_
the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). Ilv1MINENT:
The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a General Emergency.
101 11 SYSTEM MALFUNCTION ICS/EALS 102 SU1 ECL: Unusual Event Initiating Condition:
Loss of all offsite AC power capability to emergency buses for 15 minutes or longer. Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
SUI.I Loss of ALL offsite AC power capability to 1(2)A-05 and 1(2)A-06 for 15 minutes or longer. Definitions:
None Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of safety of the plant. For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it. Note: with respect to this EAL, "Station Blackout is Unit 1(2) specific." Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power. Escalation of the emergency classification level would be via IC SAL 103 SU3 ECL: Unusual Event Initiating Condition:
UNPLANNED loss of Control Room indications for 15 minutes or longer. Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
SU3.l An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.
* Reactor Power
* RCS I Pressurizer Level
* RCS I Pressurizer Pressure
* Core Exit/ RCS Temperature
* Level in at least one steam generator
* Steam Generator Auxiliary Feed Water Flow Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related. UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).
For example, the reactor power
* level cannot be determined from any analog, digital and recorder source within the Control Room. 104 An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.
The event would be reported if it significantly impaired the capability to perform emergency assessments.
In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.
In addition, if all indication sources for one or mote of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level would be via IC SA3. 105 SU4 ECL: Unusual Event Initiating Condition:
Reactor coolant activity greater than Technical Specification allowable limits. Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: SU4.1 SU4.2 Failed Fuel Monitor 1(2)RE-109 reading greater than 750 mR/hr. Sample analysis indicates tha(a RCS Specific Activity value is greater than an allowable limit specified in Technical Specifications as indicated by ANY of the following conditions:
: a. Dose Equivalent I-131 greater than 50 &#xb5;Ci/gm OR b. Dose Equivalent I-131 greater than 0.5 &#xb5;Ci/gm but less than or equal to 50 &#xb5;Ci/gm for greater than 48 hours OR c. Dose Equivalent Xe-133 greater than 300 &#xb5;Ci/gm for greater than 48 hours Definitions:
None Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specification 3.4.16 for longer than the allowed completion time. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. Escalation of the emergency classification level would be via ICs FAl or the Recognition Category R ICs. 106 
.ECL: Unusual Event Initiating Condition:
RCS leakage for 15 minutes or longer. Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: SUS Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
SUS.I RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer. SU5.2 RCS identified leakage greater than 25 gpm for 15 minutes or longer. SU5.3 Leakage from the RCS to a location outside containment, or Steam Generator tube leakage, greater than 25 gpm for 15 minutes or longer. Definitions:
UNISOLABLE:
An open or breached system line that cannot be isolated, remotely or locally. Basis: This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant. EAL SU5.1 and EAL SU5.2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications).
EAL SU5.3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus apply to leakage into the containment, a secondary-side system ( e.g., steam generator tube leakage) or a location outside of containment.
The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications.
Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation).
EAL SU5.1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. 107 The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification.
For PBNP, an emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected ( e.g., a relief valve sticks open and the line flow cannot be isolated).
The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.
* Escalation of the emergency classification level would be via ICs of Recognition Category R or F. 108 ECL: Unusual Event Initiating Condition:
Automatic or manual trip fails to shutdown the reactor. Operating Mode Applicability:
1 Emergency Action Levels: SU6 Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
SU6.l SU6.2 a. An automatic trip did not shutdown the reactor. AND b. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor. a. A manual trip did not shutdown the reactor. AND b. EITHER of the following:
: 1. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor. OR 2. A subsequent automatic trip is successful in shutting down the reactor. Definitions:
None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. NOTE: For PBNP, the phrase "at the reactor control consoles" means the reactor trip pushbuttons on the following panels:
* Unit 1 on panels 1 C04 and CO 1
* Unit 2 on panels 2C04 and CO2 109 Following the failure on an automatic reactor trip, operators will promptly initiate manual actions in the Control Room to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A.manual action in the Control Room is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core ( e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies.
Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles".
The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC F Al. Absent the plant conditions needed to meet either IC SA6 or F Al, an Unusual Event declaration is appropriate for this event. A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.
* If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.
* If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.
110 ECL: Unusual Event Initiating Condition:
Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: SU7.1 Loss of ALL of the following onsite communication methods:
* Plant Public Address System (Gai-Tronics)
* Commercial Phones
* PBXPhones
* Security Radio
* Portable Radios SU7 SU7.2 Loss of ALL of the following offsite response organization communications methods:
* Nuclear Accident Reporting System (NARS)
* Commercial Phones
* PBXPhones
* Satellite Phones
* Manitowoc County Sheriffs Department Radio SU7.3 Loss of ALL of the following NRC communications methods:
* FTS Phone System
* Commercial Phones
* PBXPhones
* Satellite Phones Definitions:
None Basis: This IC addresses a significant loss of on-site or off site communications capabilities.
While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to offsite response organizations and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). EAL SU7.1 addresses a total loss of the communications methods used in support of routine plant operations.
EAL SU7.2 addresses a total loss of the communications methods used to notify all offsite response organizations of an emergency declaration.
The offsite_response organizations referred to here are the State of Wisconsin, Manitowoc County, and Kewaunee County. EAL SU7 .3 addresses a total loss of the communications methods used to notify the NRC of an -emergency declaration.
111 SUS ECL: Unusual Event Initiating Condition:
Failure to isolate containment or loss of containment pressure control. Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: SUS.I SU8.2 a. Failure of containment to isolate when required by an actuation signal. AND b. ALL required penetrations are not closed within 15 minutes of the actuation signal. a. Containment pressure greater than 25 psig. AND b. Less than one full train of Containment Cooling System equipment is operating per design for 15 minutes or longer. Definitions:
None Basis: This IC addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant. For EAL SU8.1, the containment isolation signal must be generated as the result on an normal/accident condition (e.g., a safety injection or high containment pressure);
a failure resulting from testing or maintenance does not warrant classification.
The determination of containment and penetration status -isolated or not isolated -should be made in accordanc,e with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
EAL SU8.2 addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. During a design basis accident, a minimum of two containment accident fan cooler units with their accident fans running and one containment spray train are required to maintain the containment peak pressure and temperature below the design limits. Each containment spray train is a containment spray pump, spray header, nozzles, valves and piping. Each containment accident fan cooler unit consists of cooling coils, accident backdraft damper, accident fan, service water outlet valves, and controls necessary to ensure an operable service water flow path. 112 The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible.
The inability to start the required equipment indicates that containment heat removal/depressurization systems are either lost or performing in a degraded manner. This event would escalate to a Site Area Emergency in accordance with IC FS 1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.
113 SA1 ECL: Alert Initiating Condition:
Loss of all but one AC power source to emergency buses for 15 minutes or longer. Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that . SAl.1 15 minutes has been exceeded, or will likely be exceeded.
-a. AC power capability to 1(2)A-05 AND 1(2)A-06 is reduced to a single power source for 15 minutes or longer. AND b. Any additional single power source failure will result in a loss of ALL AC power to SAFETY SYSTEMS. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.
This IC provides an escalation path from IC SUI.
* Normal Unit 1(2) offsite power sources include: o 345 KVAC 1(2)X-03 through the 13.8 KVAC system to the LVSAT 1(2)X-04 o 345 KVAC backfed through the 19 KVAC system to the UAT 1(2)X-02
* Normal Unit 1(2) onsite power sources consist of: o em<?rgency diesel generators o . gas turbine generator o unit main turbine generator o power supplied from the opposite unit 114 An "AC power source" is a source recognized in AOPs and EOPs (including Beyond Design Basis event procedures), and capable of supplying required power to an emergency bus. Some examples of this Initiating Condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
* A loss of all off site power and loss of all emergency power sources ( e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SS 1. 115 SA3 ECL: Alert Initiating Condition:
UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: Note: The Emergency Director should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
SA3.1 a. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.
* Reactor Power
* RCS I Pressurizer Level
* RCS I Pressurizer Pressure
* Core Exit/ RCS Temperature
* Levels in at least one steam generator
* Steam Generator Auxiliary Feed Water Flow AND b. ANY of the following transient events in progress.
* Automatic or manual runback greater than 25% thermal reactor power
* Electrical load rejection greater than 25% full electrical load
* Reactor trip
* SI actuation Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. 116 As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter( s ). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.
The event would be reported if it significantly impaired the capability to perform emergency assessments.
In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.
In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level would be via ICs FS 1 or IC RS 1. 117 SA6 ECL: Alert Initiating Condition:
Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. Operating Mode Applicability:
1 Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
Emergency Action Levels: SA6.1 a. An automatic or manual trip did not shutdown the reactor. AND b. Manual actions taken at the reactor control consoles are not successful in shutting down the reactor. Definitions:
None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful.
This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS. NOTE: For PBNP, the phrase "at the reactor control consoles" means the reactor trip pushbuttons on the following panels:
* Unit 1 on panels 1C04 and COl
* Unit 2 on panels 2C04 and CO2 A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core ( e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies.
If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers).
Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control consoles." 118 The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possiple via IC FS 1. Absent the plant conditions needed to meet either IC SS6 or FS 1, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
119 SA9 ECL: Alert Initiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: Notes: SA9.1
* If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is. not warranted.
* If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.
: a. b. The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Lake level greater than or equal to +9.0 ft. (Plant elevation)
* Pump bay level less than -19.0 ft.
* Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director AND 1. Event damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode. AND 2. EITHER of the following:
* Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, or
* The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. 120 Definitions:
EXPLOSION:
A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required iflarge quantities of smoke and heat are observed.
_ SAFETY SYSTEM:_ A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second _ SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria SA9.l.b.l of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. 121 VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. High lake water level conditions that may have resulted in a plant VITAL AREA being subjected to levels beyond design limits, and thus damage may be assumed to have occurred to plant SAFETY SYSTEMS. Lake water level at +9.0 feet corresponds to the license basis flood elevation and is one foot above the Turbine Building floor elevation.
The low pump bay level threshold corresponds to the level that is calculated to correspond to the onset of cavitation of the service water pumps. Escalation of the emergency classification level would be via IC FS 1 or RS 1. 122 551 ECL: Site Area Emergency Initiating Condition:
Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer. Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: Note: The_Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. . SSl.1 Loss of ALL offsite and ALL onsite AC power to 1(2)A-05 and 1(2)A-06 for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: For the purpose of classification under this EAL, evaluation of power sources should be made on each unit individually.
This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to safety-related 4160 V AC busses. Even though a safety-related 4160 V AC bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not operable on the energized bus then the bus should not be considered operable.
If mitigative strategies establish emergency power to any bus listed in the EAL, the EAL threshold for this Initiating Condition is not met. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via I Cs RG 1, FG 1 or SG 1. 123 ECL: Site Area Emergency Initiating Condition:
Loss of all Vital DC power for 15 minutes or longer. Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
552 SS2.1 Indicated voltage is less than 115 VDC on ALL Vital DC busses D-01, D-02, D-03, and D-04 for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection o*f the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via I Cs RG 1, FG 1 or SG2. 124 556 ECL: Site Area Emergency Initiating Condition:
Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal. Operating Mode Applicability:
1 Emergency Action Levels: SS6.1 a. An automatic or manual trip did not shutdown the reactor. AND b. All manual actions to shutdown the reactor have been unsuccessful.
AND c. EITHER of the following conditions exist:
* Conditions requiring entry into Core Cooling -Red Path (CSP-C.1) are met.
* Conditions requiring entry into Heat Sink-Red Path (CSP-H.1) are met. Definitions:
None Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.
In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs.
This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor. A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalation of the emergency classification level would be via IC RG 1 or FG 1. 125 SG1 ECL: General Emergency Initiating Condition:
Prolonged loss of all offsite and all onsite AC power to emergency buses. Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: Note: The Emergency Director should declare the General Emergency promptly upon determining that 4 hours has been exceeded, or will likely be exceeded.
* SGl.1 a. Loss of ALL offsite and ALL onsite AC power to 1(2)A-05 and 1(2)A-06.
AND b. EITHER of the following:
* Restoration of at least one AC emergency bus in less than 4 hours is not likely.
* Conditions requiring entry into Core Cooling~ Red Path (CSP-C. l) are met. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: For the purpose of classification under this EAL, evaluation of power sources should be made on each unit individually.
This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers.
In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
If mitigative strategies establish emergency power to any bus listed in the EAL, the EAL threshold for this Initiating Condition is not met. The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG 1. This will allow additional time for implementation of offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses .and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.
126 The estimate for restoring at least one emergency bus should be based on a realistic appra1sal of the situation.
Mitigation actions with a low. probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public. The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core. 127 SG2. ECL: General Emergency Initiating Condition:
Loss of all AC and Vital DC power sources for 15 minutes or longer. Operating Mode Applicability:
1, 2, 3, 4 Emergency Action Levels: Note: The Emergency Director should declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
SG2.l a. Loss of ALL offsite and ALL onsite AC power to 1(2)A-05 and 1(2)A-06 for' 15 minutes or longer. AND b. Indicated voltage is less than 115 VDC on ALL Vital DC busses D-01, D-02, D-03 and D-04 for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
Basis: This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control,
* spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.
If mitigative strategies establish emergency power to any bus listed in the EAL, the EAL threshold for this Initiating Condition is not met. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met. 128 APPENDIX A -ACRONYMS AND ABBREVIATIONS AC ............................................................
: .........................................................
Alternating Current AOP ............................ , ....................................................................
Abnormal Operating Procedure ATWS ...................................................................................
Anticipated Transient Without Scram CDE ....................................................................
: .................................
Committed Dose Equivalent CFR ......................................................................................................
Code of Federal Regulations CNMT .............................
-..............................................................................................
Containment CSF .............................................................................................................
Critical Safety Function CSFST .....................
; ................................................................
Critical Safety Function Status Tree DC ..............................................................................................................................
Direct Current EAL ....................................
: ......................................................................
Emergency Action Level ECCS ............................................................................... : ............
Emergency Core Cooling System ECL ................................................................................................
Emergency Classification Level EOF ..................................................................................................
Emergency Operations Facility BOP ........................................................
." ......................................
Emergency Operating Procedure EPA .............................................................................................
Environmental Protection Agency EPG ...............................................................................................
Emergency Procedure Guideline FEMA ........................... , .................................................
Federal Emergency Management Agency GE ......................................................................................................................
General Emergency IC ........................................................................................................................
Initiating Condition ID .............................................................................................................................
Inside Diameter ISFSI ...........................................................................
Independent Spent Fuel Storage Installation Keff ....................................................................................
Effective Neutron Multiplication Factor LCO ...............................................................................................
Limiting Condition of Operation LOCA ........................................................................................................
Loss of Coolant Accident mR, mRem, mrem, mREM ............................................................
milli-Roentgen Equivalent Man MW ....................................................................................................................................
Megawatt NEI .............................................................................................................
Nuclear Energy Institute NRC ..............................................................................................
Nuclear Regulatory Commission NORAD .................................................................
North American Aerospace Defense Command NUMARC 1 .................*..............................*....*......***
Nuclear Management and Resources Council OBE .......................................................................................................
Operating Basis Earthquake OCA .............................................................................................................
Owner Controlled Area ODCM ..........................................................................................
Offsite Dose Calculation Manual PA .................
* .............................................................................................................
Protected Area P AG .......................................................................................................
Protective Action Guideline PRA/PSA ...... : .............................
Probabilistic Risk Assessment/
Probabilistic Safety Assessment PWR ........................................................................................................
Pressurized Water Reactor PSIG .................................................................................................
Pounds per Square Inch Gauge R .........................................................................................................................................
Roentgen RCS .............................................................................................................
Reactor Coolant System Rem, rem, REM ......................................................................................
Roentgen Equivalent Man RPS ............................
: ... ; ........................................................................
Reactor Protection System RPV .............................................................................................................
Reactor Pressure Vessel RVLIS ..................................................
: ...................
Reactor Vessel Level Instrumentation System 1 NUMARC was a predecessor organization of the Nuclear Energy Institute (NEI). A-1 SCBA.......
.. .. . . . . . . . . .
... . . . . .
.. . . . . . . . . . . . .
..............
.... ................
.........
Self-Contained Breathing Apparatus SG ...........................................................................................................................
Steam Generator SI ..............................................................................................................................
Safety Injection SPDS ............................................................................................
Safety Parameter Display System
* TEDE ............................................................................
,. ................
Total Effective Dose Equivalent TSC ...........................................................................................................
Technical Support Center UE ..............................................................................................................................
Unusual Event UFSAR .................................................................................
Updated Final Safety Analysis Report WOG ..................................................................................................
Westinghouse Owners Group A-2 APPENDIX B -DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatocy guidance documents.
Alert: Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA P AG exposure levels. General Emergency:
Events are in progress or have occurred which involve actual or Th1MINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Unusual Event (UE): Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases ofradioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. Site Area Emergency:
Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary. , The following are key terms necessary for overall understanding the NEI 99-01 emergency classification scheme. Emergency Action Level (EAL): A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Emergency Classification Level (ECL ): One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:
* Unusual Event (UE)
* Alert
* Site Area Emergency (SAE)
* General Emergency (GE) B-1 Fission Product Barrier Threshold:
A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Initiating Condition (IC): An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.
Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters ( e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document.
The definitions of these terms are provided below. CONFINEMENT BOUNDARY:
The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. CONTAINMENT CLOSURE: The procedurally defined conditions or actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.
EXPJ_,OSION:
A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
A release of steam ( from high energy lines 'or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.
HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE ACTION: An act toward a nuclear power plant or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.
B-2 IMMINENT:
The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. OWNER CONTROLLED AREA: This term is typically taken to mean the site property owned by" or otherwise under the control of the licensee.
* PROJECTILE:
An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal. RUPTURE(D):
The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.
SECURITY CONDITION:
Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION. SITE BOUNDARY:
That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.
UNISOLABLE:
An open or breached system line that cannot be isolated, remotely or locally. UNPLANNED:
A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.
The cause of the parameter change or event may be known or unknown. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. B-3 ATTACHMENT 3 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 286, ADOPTION OF EMERGENCY ACTION LEVEL SCHEME PURSUANT TO NEI 99-01 REVISION 6, "DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR NON-PASSIVE REACTORS" UPDATED DEVIATIONS AND DIFFERENCES MATRIX 99 pages follow PBNP DEVIATIONS AND DIFFERENCES MATRIX TABLE OF CONTENTS
 
==GENERAL COMMENT==
S ................................................................................................................................................
1 ABNORMAL RAD LEVELS/ RADIOACTIVE EFFLUENT ICS/EALS ..................................................................................
5 COLD SHUTDOWN/
REFUELING SYSTEM MALFUNCTION ICS/EALS ........................................................................
22 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ...................................................................
37 FISSION. PRODUCT BARRIER ICS/EALS ............
; .........................................................................................................
39 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ..............................................................
48 SYSTEM MALFUNCTION ICS/EALS ............................................................................................................................
64 APPENDIX A-ACRONYMS AND ABBREVIATIONS
....................................................................................................
83 APPENDIX B -DEFINITIONS
......................................................................................................................................
88 APPENDIX C-PERMANENTLY DEFUELED ICS/EALS ..................................................................................................
97 PBNP DEVIATIONS AND DIFFERENCES MATRIX
 
==GENERAL COMMENT==
S Page 1 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# GLOBAL#l References to NEI 99-01 Replaced with PBNP Difference Convert generic guidance to PBNP specific.
None GL0BAL#2 Effective date Replaced with TBD, 2018 Difference Convert generic guidance to PBNP specific.
None GL0BAL#3 Defined terms in Appendix B; Defined terms in Appendix B; Difference All defined terms in Appendix B used in the Title Case Upper Case document are in upper case (CAPs) to None indicate that the terms are defined. GL0BAL#4 BWR specific references BWR references removed Difference PBNP is a PWR None GLOBAL#S Recognition Category A-Recognition Category R-Difference PBNP implemented the optional Abnormal Radiation Abnormal Radiation designation of "R" for radiological related Levels/Radiological Effluent Levels/Radiological Effluent items to maintain continuity with previous None category and Emergency Action category and Emergency Action practice at PBNP. Levels; AU, AA, AS, and AG Levels; RU, RA, RS, and RG GL0BAL#6 Permanently Defueled Section Deleted references to Difference Not Applicable to PBNP Permanently Defueled Station None GL0BAL#7 Acknowledgments, Notice and Deleted Difference Not Applicable to PBNP Executive Summary None GL0BAL#8 Parameters or indications listed Some parameters or indications Difference Tables or bullets were created to present in EALs listed in EALs were placed in PBNP-specific information in a manner None tables or bulletized lists. familiar to and desired by scheme users. GL0BAL#9 Site specific information or Site specific information or Difference Compliance with intent of the guidance.
indication statements indication statements were replaced with PBNP information None or indications where applicable and the statement deleted. GLOBAL#lO Operating Mode Applicability lists Operating Mode Applicability lists Difference Mode numbers used for consistency with mode names (i.e., Power mode numbers (i.e., Modes 1 and PBNP procedures and training.
None Operation, Startup) 2) GLOBAL#ll Developer's Notes Developer's Notes deleted Difference Developer's notes are not reflected in the implementation of the EALs. None GL0BAL#12 Example EAL statement "Example" deleted from Difference In adopting the EAL, the "example" status statement is no longer applicable.
None GL0BAL#13 The following terms: "all, any" Consistently capitalized and Difference Capitalized and balded conditional terms in are sometimes capitalized and/or balded the following terms: "all, ICs and EALs for consistency based on user None balded in ICs and EALs any" in ICs and EALs. feedback.
GL0BAL#14 Defined terms are only listed in Defined terms are also listed as in Difference Aid to the user to present all needed APPENDIX B -DEFINITIONS separate section of each IC/EAL information within the same section of the where the defined terms are Basis document.
None used. 2 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# COVER PAGE Development of Emergency Point Beach Emergency Action Difference Changes made to adapt the generic NEI None Action Levels for Non-Passive Level Bases Document guidance to a PBNP-specific document Reactors Introduction Acknowledgments, Notice and Deleted Difference Not Applicable to PBNP None Executive Summary TOC 1. Regulatory Background
: 1. Development of Emergency Difference Title change None Action Levels TOC 1.1 Operating Reactors 1.1 Regulatory Background Difference Title change None TOC 1.2 Permanently Defueled Station Deleted section Difference Not Applicable to PBNP None TOC 1.3 Independent Spent Fuel 1.2 Independent Spent Fuel Difference Re-numbered None Storage Installation (ISFSI) Storage Installation (ISFSI) TOC 1.4 NRC Order EA-12-051 1.3 NRC Order EA-12-051 Difference Re-numbered None TOC 1.5 Applicability of Advance and Deleted section Difference Not Applicable to PBNP None Small Modular Reactor Designs TOC 3.Design ofthe NEI 99-01 3. Design of the PBNP Emergency Difference Title Change None Emergency Classification Scheme Classification Scheme TOC 3.3 NSSS Design Differences Deleted section Difference Changes made to adapt the generic NEI None guidance to a PBNP-specific document TOC 3.4 Organization and Changed to 3.3 PBNP 3.4 Difference Changes made to adapt the generic NEI None Presentation of Generic Organization and Presentation of guidance to a PBNP-specific document Information Generic Information TOC 4.0 Site-Specific Scheme 4.0 PBNP Scheme Development Difference Title change None Development TOC 4.4; 4.5; 4.6;
 
===4.8 Deleted===
sections Difference Changes made to adapt the generic NEI None guidance to a PBNP-specific document TOC 4.7 Developer and User Feedback None TOC Appendix (-Permanently Deleted section Difference Changes made to adapt the generic NEI None Defueled Station ICs/EALs guidance to a PBNP-specific document 1.1 Regulatory Background Regulatory Background Difference Changes made to adapt the generic NEI None . guidance to a PBNP-specific document and removed developer information
 
===1.2 Permanently===
 
Defueled Station Section deleted Difference Not Applicable to PBNP None 1.3 1.3 Independent Spent Fuel 1.2 Independent Spent Fuel Difference Re-numbered section and renamed E-HU1 None Storage Installation (ISFSI) Storage Installation (ISFSI) to EU1 to prevent user confusion.
1.4 1.4 NRC Order EA-12-051 1.3 NRC Order EA-12-051 Difference Re-numbered and removed wording to add None these readings once the instruments are installed (PBNP installation completed).
3 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# 1.5 Applicability to Advanced and Section deleted Difference Not Applicable to PBNP None Small Modular Reactor Designs 2 KEY TERMINOLOGY USED IN NEI KEY TERMINOLOGY USED IN NEI Difference Minor changes to reflect PBNP-specific None 99-01 99-01 implementation.
3 DESIGN OF THE NEI 99-01 DESIGN OF THE PBNP Difference Changes made to adapt the generic NEI None EMERGENCY CLASSIFICATION EMERGENCY CLASSIFICATION guidance to a PBNP-specific document SCHEME SCHEME 3.1 Assignment of Emergency Assignment of Emergency Difference Changes made to adapt the generic NEI None Classification Levels (ECLs) Classification Levels (ECLs) guidance to a PBNP-specific document, removed references to BWRs, and removed developer information.
 
===3.2 Types===
of Initiating Conditions and Types of Initiating Conditions and Verbatim None Emergency Action Levels Emergency Action Levels 3.3 Text referring to NSSS design Deleted Difference Guidance is now PBNP specific None differences for various types or nuclear plants; Developer guidance 3.4 Organization and Presentation of PBNP-Specific Organization and Difference Renumbered to 3.3, made PBNP-specific, None Generic Information Presentation of Generic and deleted developer information Information 3.5 Mode of Applicability Matrix; Deleted "Permanently Defueled" Difference Renumbered to 3.4, removed BWR Vl Typical PWR Operating Modes section of matrix; replaced information, removed permanently Typical PWR Operating Modes defueled, and inserted PBNP Operating with PBNP Operating Modes Modes to com ply with the intent of the document.
Global comment #5 4 Site Specific Scheme PBNP Scheme Development Difference Updated to reflect PBNP specific scheme None Development Guidance development process. 5 GUIDANCE ON MAKING GUIDANCE ON MAKING Difference Added text from Section IV.H. 7 of V2 EMERGENCY CLASSIFICATIONS EMERGENCY CLASSIFICATIONS NSIR/DPR-ISG-01 explaining how to treat concurrent time periods when making an emergency declaration.
Information was added to address a frequently asked question by the PBNP operators.
6 -11 Recognition Category IC/EAL removed Difference Matrixes were intended for use by EAL None Matrixes developers.
Not included in licensee scheme. 4 PBNP DEVIATIONS AND DIFFERENCES MATRIX ABNORMAL RAD LEVELS/ RADIOACTIVE EFFLUENT ICS/EALS 5 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
AUl RU1 Difference Global Comment #5 None Initiating Condition:
Release of Release of gaseous or liquid Difference Global Comment #9 None gaseous or liquid radioactivity radioactivity greater than 2 times greater than 2 times the (site-the ODCM limits for 60 minutes .... => specific effluent release or longer. <C controlling document) limits for 60 minutes or longer. Operating Mode of Applicability:
Operating Mode of Applicability:
Verbatim None All All 6 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# (1) Reading on ANY effluent (1) Reading on ANY of the Difference See Global Comments #8, 9, 12, & 13. V3 &V4 radiation monitor greater following effluent radiation than 2 times the (site-monitors greater than the Reworded EAL statement to remove specific effluent release reading shown for 60 operator confusion as to whether they controlling document) minutes or longer: needed to multiply the values of the limits for 60 minutes or Monitor Reading following table by 2 or if the value provided longer: (site-specific 1(2)-RE-307 CTMNT Purge 1.4E-2 already was 2X. Wording now matches monitor list and threshold Exhaust Mid-Range Gas with only &#xb5;Ci/cc wording of RSl and RG1 allowing for easier values corresponding to 2 containment purge in operation operator progression through the EALs. times the controlling 2-RE-30S CTMNT Purge Exhaust Low Range Gas with both purge 9.4E-3 document limits) and GS building ventilation in &#xb5;Ci/cc operation 2-RE-307 CTMNT Purge Exhaust Mid-Range Gas with both purge 9.4E-3 and GS building ventilation in &#xb5;Ci/cc ,.-._ operation
...: = 2-RE-307 CTMNT Purge Exhaust 0 2.8E-2 Mid-Range Gas with only GS '-' &#xb5;Ci/cc building ventilation in operation 2-RE-307 CTMNT Purge Exhaust 1.0E+l Mid-Range Gas with only forced &#xb5;Ci/cc vent of containment 2-RE-309 CTMNT Purge Exhaust 1.0E+l High Range Gas with only forced &#xb5;Ci/cc vent of containment RE-315 AB Exhaust Low Range 5.4E-3 Gas &#xb5;Ci/cc RE-317 AB Exhaust Mid-Range 5.4E-3 Gas &#xb5;Ci/cc RE-325 Drumming Area Exhaust 8.4E-3 Low Range Gas &#xb5;Ci/cc RE-327 Drumming Area Exhaust 8.4E-3 Mid-Range Gas &#xb5;Ci/cc 1(2) RE-229 Service Water 2.3E-3 Overboard
&#xb5;Ci/cc 7 _____ J PBNP DEVIATIONS AND DIFFERENCES MATRIX ..*.. Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# (2) Reading on ANY effluent (2) Reading on ANY effluent Difference Global Comment #13 None radiation monitor greater radiation monitor greater than 2 times the alarm than 2 times the alarm setpoint established by a setpoint established by a current radioactivity current radioactivity discharge permit for 60 discharge permit for 60 minutes or longer. minutes or longer. {3) Sample analysis for a {3) Sample analysis for a gaseous Difference Global Comment #9 None -gaseous or liquid release or liquid release indicates a ...: indicates a concentration concentration or release rate r: 0 greater than 2 times the or release rate greater ..-1 than 2 times the (site-ODCM limits for 60 minutes ::::, <( specific effluent release or longer. controlling document) limits for 60 minutes or longer. Intent and meaning of the EALs are not altered. 8 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
AU2 RU2 Difference Global Comment #5 & 14 None Initiating Condition:
UNPLANNED UNPLANNED loss of water level Verbatim None loss of water level above above irradiated fuel. irradiated fuel. Operating Mode of Applicability:
Operating Mode of Applicability:
Verbatim None All All (1) a. UNPLANNED water level (1) a UNPLANNED water level Difference Global Comment #9, 12 & 13 None drop in the REFUELING drop in the REFUELING PATHWAY as indicated by PATHWAY as indicated by ANY of the following:
ANY of the following: (site-specific level
* Spent fuel pool low indications).
water level alarm
* Visual observation AND AND b. UNPLANNED increase in b. UNPLANNED rise in area Difference Global Comments #9 & 13 None area radiation levels as radiation levels as N indicated by ANY of the indicated by ANY of the ::::, <C following radiation following radiation monitors.
monitors. (site-specific list of area
* RE-105 SFP Area Low radiation monitors)
Range Radiation Monitor
* RE-135 SFP Area High Range Radiation Monitor
* 1(2) RE-102 El. 66' CONTAINMENT Low Range Monitor Intent and meaning of the EALs are not altered. 9 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
AA1 RA1 Difference Global Comment #5 & 14 None lnitiati11g condition:
Release of Release of gaseous or liquid Verbatim None gaseous or liquid radioactivity radioactivity resulting in offsite ,-1 resulting in offsite dose greater dose greater than 10 mrem TEDE <C than 10 mrem TEDE or 50 mrem or 50 mrem thyroid CDE. <C thyroid CDE. Operating Mode of Applicability:
Operating Mode of Applicability:
Verbatim None All All 10 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section I NEI 99-01 Rev. 6 PBNP Change Justification Validation
# (1) Reading on ANY of the (1) Reading on ANY of the Difference Global Comment #8, 9, 12 & 13 vs following radiation following radiation monitors monitors greater than the greater than the reading reading shown for 15 shown for 15 minutes or minutes or longer: longer: (site-specific monitor list and Monitor Reading threshold values) 1{2)-RE-307 CTMNT Purge 6.0E+O Exhaust Mid-Range Gas with &#xb5;Ci/cc only containment purge in operation 1(2)-RE-309 CTMNT Purge 6.0E+o Exhaust High Range Gas with &#xb5;Ci/cc only containment purge in -operation
...; C 2-RE-307 CTMNT Purge 4.0E+O 0 Exhaust Mid-Range Gas with &#xb5;Ci/cc .-l both purge and GS building <( ventilation in operation
<( 2-RE-309 CTMNT Purge 4.0E+O Exhaust High Range Gas with &#xb5;Ci/cc both purge and GS building ventilation in operation 2-RE-309 CTMNT Purge 1.2E+1 Exhaust High Range Gas with &#xb5;Ci/cc only GS building ventilation in operation RE-317 AB Exhaust Mid-Range 1.0E+O Gas &#xb5;Ci/cc RE-319 AB Exhaust High Range 1.0E+O Gas &#xb5;Ci/cc RE-327 Drumming Area 1.6E+O Exhaust Mid-Range Gas &#xb5;Ci/cc 11 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #9 None actual meteorology meteorology indicates doses indicates doses greater greater than 10 mrem TEDE than 10 mrem TEDE or SO or SO mrem thyroid CDE at or mrem thyroid CDE at or beyond the SITE BOUNDARY.
beyond (site-specific dose receptor point). (3) Analysis of a liquid (3) Analysis of a liquid effluent Difference Global Comment #9 effluent sample indicates sample indicates a a concentration or release concentration or release rate rate that would result in that would result in doses doses greater than 10 greater than 10 mrem TEDE mrem TEDE or SO mrem or SO mrem thyroid CDE at or thyroid CDE at or beyond beyond the SITE BOUNDARY (site-specific dose for one hour of exposure.
receptor point) for one hour of exposure.
(4) Field survey results (4) Field survey results indicate Difference Global Comment #9 indicate EITHER of the EITHER of the following at or following at or beyond beyond the SITE BOUNDARY: (site-specific dose
* Closed window dose receptor point): rates greater than 10 -* Closed window dose mR/hr expected to ,.; rates greater than 10 continue for 60 minutes C 0 ..:!. mR/hr expected to or longer. .-I continue for 60 Analyses offield survey
* minutes or longer. samples indicate thyroid
* Analyses of field survey CDE greater than SO samples indicate mrem for one hour of thyroid CDE greater inhalation.
than SO mrem for one hour of inhalation.
Intent and meaning of the EALs are not altered. 12 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
AA2 RA2 Difference Global Comment #5 & 14 None Initiating Condition: Significant Significant lowering of water Verbatim None lowering of water level above, or level above, or damage to, damage to, irradiated fuel. irradiated fuel. Operating Mode of Applicability:
Operating Mode of Applicability:
Verbatim None All All (1) Uncovery of irradiated fuel (1) Uncovery of irradiated fuel in Verbatim None in the REFUELING the REFUELING PATHWAY. PATHWAY. (2) Damage to irradiated fuel (2) Damage to irradiated fuel Difference Global Comment #8, 9, 12 & 13 V6 resulting in a release of resulting in a release of radioactivity from the fuel radioactivity from the fuel as as indicated by ANY of the indicated by reading on ANY following radiation of the following radiation monitors:
monitors greater than the N reading shown: (site-specific listing of radiation monitors, and the associated Monitor Reading readings, setpoints and/or RE-105 SFP Area Low 4 R/hr Range Radiation alarms) Monitor 1(2)-RE-126 7 R/hr Containment High Radiation Monitor 1(2)-RE-127 7 R/hr Containment High Radiation Monitor 1(2)-RE-128 7 R/hr Containment High Radiation Monitor (3) Lowering of spent fuel pool (3) Lowering of spent Difference Global Comment #9 V7 level to (site-specific Level fuel pool level to 49 2 value). [See Developer ft. 0 in. Intent and meaning of the EALs are not Notes altered. 13 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99~01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
AA3 RA3 Difference Global Comment #5 & 14 None Initiating Condition:
Radiation Radiation levels that impede access Verbatim None levels that impede access to to equipment necessary for normal equipment necessary for normal plant operations, cooldown or plant operations, cooldown or shutdown.
shutdown.
Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) Dose rate greater than 15 (1) Dose rate greater than 15 Difference Global Comment #9, 12 & 13 None mR/hr in ANY of the mR/hr in ANY of the following following areas: areas:
* Control Room
* Control Room (RE-101)
* Central Alarm Station
* Central Alarm Station AND * (other site-specific Secondary Alarm Station areas/rooms) (by survey) (2) An UNPLANNED event (2) An UNPLANNED event results Difference Global Comment #8, 9 & 13 V8 results in radiation levels in radiation levels that prohibit m that prohibit or impede or impede access to ANY of the <( access to any of the following plant rooms or areas: <( following plant rooms or areas: Table R-2 SAFE OPS, S/D, C/D AREAS (site-specific list of plant rooms Area/Building MODE or areas with entry-related mode Ul VCT Area 3/4/5 applicability identified)
U2 VCT Area 3/4/ 5 Ul Primary Sample area 3 U2 Primary Sample area 3 CCW HX Room 4/5 C-59 area 3/4/5 Pipeway 2, 8 ft. Elev. 3/4 Pipeway 3, 8 ft. Elev. 3/4 1/2832 MCC Area 4 Intent and meaning of the EALs are not altered. 14 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
AS1 RS1 Difference Global Comment #5 & 14 None Initiating Condition:
Release of Release of gaseous radioactivity Verbatim None gaseous radioactivity resulting in resulting in offsite dose greater .... offsite dose greater than 100 than 100 mrem TEDE or 500 V'l <( mrem TEDE or 500 mrem thyroid mrem thyroid COE. COE. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None 15 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section . NEI 99-01 Rev .. 6 PBNP Change Justification Validation
# (1) Reading on ANY of the (1) Reading on ANY of the Difference Global Comment #8, 9, 12 & 13 vs following radiation following radiation monitors monitors greater than the greater than the reading reading shown for 15 shown for 15 minutes or minutes or longer: longer: (site-specific monitor list and Monitor Reading threshold values) 1(2)-RE-307 CTMNT Purge 6.0E+l Exhaust Mid-Range Gas with &#xb5;Ci/cc only containment purge in operation 1(2)-RE-309 CTMNT Purge 6.0E+l Exhaust High Range Gas with &#xb5;Ci/cc only containment purge in operation 2-RE-307 CTMNT Purge Exhaust 4.0E+l Mid-Range Gas with both purge &#xb5;Ci/cc and GS building ventilation in operation  RE-309 CTMNT Purge Exhaust 4.0E+l ..; High Range Gas with both &#xb5;Ci/cc C: purge and GS building 0 ventilation in operation
'""' V) 2-RE-309 CTMNT Purge Exhaust 1.2E+2 <( High Range Gas with only GS &#xb5;Ci/cc building ventilation in operation RE-317 AB Exhaust Mid-Range 1.0E+l Gas &#xb5;Ci/cc RE-319 AB Exhaust High Range 1.0E+l Gas &#xb5;Ci/cc RE-327 Drumming Area Exhaust 1.6E+l Mid-Range Gas &#xb5;Ci/cc 16 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #3 & 9 None actual meteorology meteorology indicates doses indicates doses greater greater than 100 mrem TEDE than 100 mrem TEDE or or 500 mrem thyroid COE at 500 mrem thyroid COE at or beyond the SITE or beyond (site-specific BOUNDARY.
dose receptor point). (3) Field survey results {3) Field survey results indicate Difference Global Comment #3 & 9 None indicate EITHER of the EITHER of the following at or following at or beyond beyond the SITE BOUNDARY: (site-specific dose receptor
* Closed window dose point): rates greater than 100
* Closed window dose mR/hr expected to rates greater than 100 continue for 60 mR/hr expected to minutes or longer. -continue for 60 minutes
* Analyses offield survey .... or longer. samples indicate C: 0 thyroid COE greater
* Analyses of field survey ...... samples indicate thyroid than 500 mrem for one V) <( COE greater than 500 hour of inhalation.
mrem for one hour of inhalation.
Intent and meaning of the EALs are not altered. 17 PBNP DEVIATIONS AND DIFFERENCES MATRIX ,, * *Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
AS2 RS2 Difference Global Comment #5 None Initiating Condition:
Spent fuel Spent fuel pool level at 40 ft. 8 in. Difference Global Comment #9 V7 pool level at (site-specific Level 3 description).
Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None N (1) Lowering of spent fuel pool V, (1) Lowering of spent fuel pool Difference Global Comment #9 & 12 V7 <( level to (site-specific Level level to 40 ft. 8 in. 3 value). Intent and meaning of the EALs are not , altered. 18 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
AGl RGl Difference Global Comment #5 & 14 None Initiating Condition:
Release of Release of gaseous radioactivity Verbatim None gaseous radioactivity resulting in resulting in offsite dose greater offsite dose greater than 1,000 than 1,000 mrem TEDE or 5,000 mrem TEDE or 5,000 mrem mrem thyroid CDE. thyroid CDE. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) Reading on ANY of the (1) Reading on ANY of the Difference Global Comment #8, 9, 12 & 13 vs following radiation following radiation monitors monitors greater than the greater than the reading reading shown for 15 shown for 15 minutes or minutes or longer: longer: .-4 (site-specific monitor list and Monitor Reading l!' threshold values) <( 1(2)-RE-309 CTMNT Purge Exhaust High Range Gas with 6.0E+2 only containment purge in &#xb5;Ci/cc operation 2-RE-309 CTMNT Purge Exhaust High Range Gas with 4.0E+2 both purge and GS building &#xb5;Ci/cc ventilation in operation 2-RE-309 CTMNT Purge Exhaust High Range Gas with 1.2E+3 only GS building ventilation in &#xb5;Ci/cc operation RE-317 AB Exhaust Mid-Range 1.0E+2 Gas &#xb5;Ci/cc RE-319 AB Exhaust High Range 1.0E+2 Gas &#xb5;Ci/cc 19 PBNP DEVIATIONS AND DIFFERENCES MATRIX
* Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# (2) Dose assessment using (2) Dose assessment using actual Difference Global Comment #3 & 9 None actual meteorology meteorology indicates doses indicates doses greater greater than 1,000 mrem than 1,000 mrem TEDE or TEDE or 5,000 mrem thyroid 5,000 mrem thyroid CDE at CDE at or beyond the SITE or beyond (site-specific BOUNDARY.
dose receptor point). (3) Field survey results (3) Field survey results indicate Difference Global Comment #3 & 9 None indicate EITHER of the EITHER of the following at or following at or beyond beyond the SITE BOUNDARY: (site-specific dose receptor
* Closed window dose point): rates greater than 1,000 -* Closed window dose mR/hr expected to .... rates greater than 1,000 continue for 60 minutes C 0 mR/hr expected to or longer. ..-l continue for 60 minutes Analyses of field survey I!,
* c( or longer. samples indicate thyroid
* Analyses of field survey CDE greater than 5,000 samples indicate thyroid mrem for one hour of CDE greater than 5,000 inhalation.
mrem for one hour of inhalation.
Intent and meaning of the EALs are not altered. 20 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
AG2 RG2 Difference Global Comment #5 None Initiating Condition:
Spent fuel Spent fuel pool level cannot be Difference Global Comment #9 V7 pool level cannot be restored to restored to at least 40 ft. 8 in. for at least (site-specific Level 3 60 minutes or longer. description) for 60 minutes or longer. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None N (1) Spent fuel pool level cannot (1) Spent fuel pool level cannot Difference Global Comment #9 & 12 V7 I.!' c:( be restored to at least (site-be restored to at least 40 ft. 8 specific Level 3 value) for 60 in. for 60 minutes or longer. minutes or longer. Intent and meaning of the EALs are not altered. 21 PBNP DEVIATIONS AND DIFFERENCES MATRIX COLD SHUTDOWN/
REFUELING SYSTEM MALFUNCTION ICS/EALS 22 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
CUl CUl Verbatim Global Comment #14 None Initiating Condition:
UNPLANNED UNPLANNED loss of reactor Difference Global Comment #4 None loss of (reactor vessel/RCS
[PWR] vessel/RCS inventory for 15 or RPV [BWR]) inventory for 15 minutes or longer minutes or longer. Operating Mode Applicability:
Operating Mode Applicability:
5, Difference Global Comment #10 None Cold Shutdown, Refueling 6 (1) UNPLANNED loss of (1) UNPLANNED loss of reactor Difference Global Comment #4 & 12 None reactor coolant results in coolant results in reactor (reactor vessel/RCS
[PWR] vessel/RCS level less than a or RPV [BWR]) level less required lower limit for 15 than a required lower limit minutes or longer for 15 minutes or longer. (2) a. (Reactor vessel/RCS
[PWR] (2) a. Reactor vessel/RCS level Difference Global Comment #4 None or RPV [BWR]) level cannot cannot be monitored.
be monitored.
AND AND .... ::> b. UNPLANNED increase in b. UNPLANNED increase in Difference Global Comment #9 V9 u (site-specific sump and/or Containment Sump A OR tank) levels. Waste Holdup Tank levels. Intent and meaning of the EALs are not altered. 23 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
CU2 CU2 Verbatim Global Comment #14 None Initiating Condition:
Loss of all Loss of all but one AC power Verbatim None but one AC power source to source to emergency buses for 15 emergency buses for 15 minutes minutes or longer. or longer. Operating Mode Applicability:
Operating Mode Applicability:
5, Difference Global Comment #10 None Cold Shutdown, Refueling, 6, Defueled Defueled (1) a. AC power capability to (1) a. AC power capability to Difference Global Comment #9 & 12 VlO (site-specific emergency 1(2)-A05 and 1(2)-A06 is N buses) is reduced to a reduced to a single power :) u single power source for 15 source for 15 minutes or minutes or longer. longer. AND AND b. Any additional single b. Any additional single power source failure will power source failure will result in loss of all AC result in loss of all AC power to SAFETY SYSTEMS. power to SAFETY SYSTEMS. Intent and meaning of the EALs are not altered. 24 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
CU3 CU3 Verbatim Global Comment #14 None Initiating Condition:
UNPLANNED UNPLANNED increase in RCS Verbatim None increase in RCS temperature.
tern peratu re. Operating Mode Applicability:
Operating Mode Applicability:
5, Difference Global Comment #10 None Cold Shutdown, Refueling 6 (1) UNPLANNED increase in (1) UNPLANNED increase in RCS Difference Global Comment #9 & 12 Vl RCS temperature to temperature to greater than greater than (site-specific 200&deg;F Technical Specification cold shutdown temperature limit). (2) Loss of ALL RCS (2) Loss of ALL RCS temperature Difference Global Comment #4 & 13 None temperature and (reactor and reactor vessel/RCS level m vessel/RCS
[PWR] or RPV indication for 15 minutes or :> [BWR]) level indication for longer u 15 minutes or longer. Intent and meaning of the EALs are not altered. 25 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change I Justification Validation
# ,, Recognition Category: CU4 CU4 Verbatim Global Comment #14 None Initiating Condition:
Loss of Vital Loss of Vital DC power for 15 Verbatim None DC power for 15 minutes or minutes or longer. longer. Operating Mode Applicability:
Operating Mode Applicability:
5, Difference Global Comment #10 None Cold Shutdown, Refueling 6 (1) Indicated voltage is less (1) Indicated voltage is less than Difference Global Comment #9 & 12 Vll 'Of" ::::, than (site-specific bus 115VDC on required Vital DC u voltage value) on required buses D-01, D-02, D-03 or D-Vital DC buses for 15 04 for 15 minutes or longer minutes or longer. Intent and meaning of the EALs are not altered. 26 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP I Change Justification Validation
# Recognition Category:
CUS cus Verbatim None Initiating Condition:
Loss of all Loss of all onsite or offsite Verbatim None onsite or offsite communications communications capabilities.
capabilities.
Operating Mode Applicability:
Operating Mode Applicability:
5, Difference Global Comment #10 None Cold Shutdown, Refueling, 6, Defueled Defueled (1) Loss of ALL of the following (1) Loss of ALL of the following Difference Global Comment #9, 12 & 13 None onsite communication onsite communication methods: methods: (site-specific list of
* Plant Public Address communications methods) System (Gai-Tronics)
* Commercial Phones
* PBX Phones
* Security Radio
* Portable Radio (2) Loss of ALL of the following (2) Loss of ALL of the following Difference Global Comment #9, 12 & 13 None ORO communications offsite response organization Ll'I methods: communications methods: ::::> u (site-specific list of Nuclear Accident
* communications methods) Reporting System (NARS)
* Commercial Phones
* PBX Phones
* Satellite Phones
* Manitowoc County Sheriff's Department Radio (3) Loss of ALL of the following (3) Loss of ALL of the following Difference Global Comment #9, 12 & 13 None NRC communications NRC communications methods: methods: (site-specific list of
* FTS Phone System communications methods)
* Commercial Phones
* PBX Phones
* Satellite Phones Intent and meaning of the EALs are not altered. 27 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
CAl CAl Verbatim Global Comment #14 None Initiating Condition:
Loss of Loss of reactor vessel/RCS Difference Global Comment #4 None (reactor vessel/RCS
[PWR] or RPV inventory.
[BWR]) inventory.
Operating Mode Applicability:
Operating Mode Applicability:
5, Difference Global Comment #10 None Cold Shutdown, Refueling 6 (1) Loss of (reactor vessel/RCS (1) Loss of reactor vessel/RCS Difference Global Comment #4, 9 & 12 V12 [PWR] or RPV [BWR]) inventory as indicated by inventory as indicated by level less than 16% on LI-level less than (site-specific 447 / Ll-447 A level). (2) a. (Reactor vessel/RCS
[PWR] (2) a. Reactor vessel/RCS level Difference Global Comment #4 None or RPV [BWR]) level cannot cannot be monitored for .-l <( be monitored for 15 15 minutes or longer u minutes or longer AND AND Difference Global Comment #4, 9 & 13 V9 b. UNPLANNED increase in b. UNPLANNED increase in (site-specific sump and/or Containment Sump A OR tank) levels due to a loss of Waste Holdup Tank levels (reactor vessel/RCS
[PWR] due to a loss of reactor or RPV [BWR]) inventory.
vessel/RCS inventory.
Intent and meaning of the EALs are not altered. 28 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation#
Recognition Category:
CA2 CA2 Verbatim Global Comment #14 None Initiating Condition:
Loss of all Loss of all offsite and all onsite Verbatim None offsite and all onsite AC power to AC power to emergency buses for emergency buses for 15 minutes 15 minutes or longer. or longer. Operating Mode Applicability:
Operating Mode Applicability:
5, Difference Global Comment #10 None N c( Cold Shutdown, Refueling, 6, Defueled u Defueled (1) Loss of ALL offsite and ALL (1) Loss of ALL offsite and ALL Difference Global Comment #9, 12 & 13 VlO onsite AC Power to (site-onsite AC Power to 1(2)-A05 specific emergency buses) and 1(2)-A06 for 15 minutes for 15 minutes or longer. or longer. Intent and meaning of the EALs are not altered. 29 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
CA3 CA3 Verbatim Global Comment #14 None Initiating Condition:
Inability to Inability to maintain the plant in Verbatim None maintain the plant in cold cold shutdown.
shutdown.
Operating Mode Applicability:
Operating Mode Applicability:
5, Difference Global Comment #10 None Cold Shutdown, Refueling 6 (1) UNPLANNED increase in (1) UNPLANNED increase in RCS Difference Global Comment #9 & 12 Vl RCS temperature to temperature to greater than greater than (site-specific 200&deg;F for greater than the Technical Specification duration specified in the cold shutdown following table: temperature limit) for greater than the duration specified in the following table. Table: RCS Heat-up Duration Threi RCS Heat-up Duration Threshc Difference Global Comment #9 None Containment RCS Status Containment Closure m RCS Status Closure Status Status Replaced "RCS heat removal system" with <C u Intact (but not at Intact (but not at "RHR" to reflect site-specific nomenclature reduced inventory Not applicable reduced inventory)
Not Applicable familiar to the operators.
[PWR]) Not intact (or at reduced inventory)
Established Not intact (or at reduced Established inventory
[PWR]) Not Established Not Established
* If RHR is in operation within this time frame temperature is being reduced, the EAL is not a
* If an RCS heat removal system is in operatic frame and RCS temperature is being reduce applicable.
(2) UNPLANNED RCS pressure (2) UNPLANNED RCS pressure Difference Global Comment #4 & 9 V13 increase greater than (site-increase greater than 25 psig. specific pressure reading). (This EAL does not apply (This EAL does not apply during water-solid plant during water-solid plant conditions.)
conditions.
[PWR]) Intent and meaning of the EAL~ are not altered. 30 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
CA6 CA6 Verbatim Global Comment #14 None Initiating Condition: Hazardous Hazardous event affecting a Verbatim None event affecting a SAFETY SYSTEM SAFETY SYSTEM needed for the needed for the current operating current operating mode. mode. Operating Mode Applicability:
Operating Mode Applicability:
5, Difference Global Comment #10 None Cold Shutdown, Refueling 6 (1) a. The occurrence of ANY of (1) a. The occurrence of ANY of Difference Global Comment #9, 12 & 13 V14 the following hazardous the following hazardous events: events:
* Seismic event
* Seismic event (earthquake) (earthquake)
* Internal or external
* Internal or external flooding event flooding event
* High winds or tornado
* High winds or tornado strike strike
* FIRE
* FIRE IJ)
* EXPLOSION
* EXPLOSION
<( * (site specific hazards)
* Lakelevelgreaterthan u
* Other events with or equal to 9.0 ft. (Plant similar hazard elevation) characteristics as
* Other events with determined by the Shift similar hazard Manager characteristics as determined by the Shift Manager AND AND VlS b. EITHER of the following:
: b. 1. Event damage has Deviation Adopted the revised EAL wording provided 1. Event damage has caused indications of in proposed EAL FAQ 2016-02. caused indications of degraded degraded performance performance in one in at least one train of a train of a SAFETY SAFETY SYSTEM needed SYSTEM needed for for the current the current operating operating mode. mode. AND 31 PBNP DEVIATIONS AND DIFFERENCES MATRIX ~ection NEI 99-01 Rev. 6 PBNP Change Justification Validation
# OR 2. EITHER of the Deviation Adopted the revised EAL wording provided V15 1. The event has caused following:
in proposed EAL FAQ 2016-02. VISIBLE DAMAGE to a
* Event damage has SAFETY SYSTEM caused indications component or structure of degraded needed for the current performance to a operating mode. second train of the SAFETY SYSTEM Intent and meaning of the EALs are not needed for the altered. current operating mode, or
* The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. 32 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
CSl CSl Verbatim Global Comment #14 None Initiating Condition:
Loss of Loss of reactor vessel/RCS Difference Global Comment #4 None (reactor vessel/RCS
[PWR] or RPV inventory affecting core decay [BWR]) inventory affecting core heat removal capability.
decay heat removal capability.
Operating Mode Applicability:
Operating Mode Applicability:
5, Difference Global Comment #10 None Cold Shutdown, Refueling 6 (1) a. CONTAINMENT CLOSURE Not used Difference Global Comment #9 & 12 None not established.
PBNP design and operation of water level .-I AND instrumentation is such that this level value II') b. (Reactor vessel/RCS
[PWR] cannot be accurately determined during u or RPV [BWR]) level less Cold Shutdown or Refueling modes, than (site-specific level). therefore this EAL was not included.
(1) a. CONTAINMENT CLOSURE Not used Difference Global Comment #4 & 9 None established.
PBN P design and operation of water level AND instrumentation is such that this level value b. (Reactor vessel/RCS
[PWR] cannot be accurately determined during or RPV [BWR]) level less Cold Shutdown or Refueling modes, than (site-specific level). therefore this EAL was not included.
33 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Jus*tification Validation
# (3) a. (Reactor vessel/RCS
[PWR] (1) a. Reactor vessel/RCS level Difference Global Comment #4 None or RPV [BWR]) level cannot cannot be monitored for be monitored for 30 30 minutes or longer. minutes or longer. AND AND b. Core uncovery is indicated
: b. Core uncovery is indicated Difference Global Comment #9 &13 V9 & V16 by ANY of the following:
by ANY of the following:
* (Site-specific radiation
* Containment High -monitor) reading greater Radiation Monitor .... than (site-specific value) (1(2)-RE-126, RE-127, C 0 or RE-128) reading
* Erratic source range .-I monitor indication greater than 100 R/hr Vl u [PWR]
* Erratic source range
* UNPLANNED increase in monitor indication (site-specific sump
* UNPLANNED increase and/or tank) levels of in Containment Sump sufficient magnitude to A OR Waste Holdup Intent and meaning of the EALs are not indicate core uncovery Tank levels of altered. * (Other site-specific sufficient magnitude indications) to indicate core uncovery 34 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
CGl CGl Verbatim Global Comment #14 None Initiating Condition:
Loss of Loss of reactor vessel/RCS Difference Global Comment #4 None (reactor vessel/RCS
[PWR] or RPV inventory affecting fuel clad [BWR]) inventory affecting fuel integrity with containment clad integrity with containment challenged.
challenged.
Operating Mode Applicability:
Operating Mode Applicability:
5, Difference Global Comment #10 None Cold Shutdown, Refueling 6 (1) a. (Reactor vessel/RCS
[PWR] b. Not used Difference Global Comment #4, 9, 12 & 13 None .-i or RPV [BWRJ) level less PBNP design and operation of water level than (site-specific level) for instrumentation is such that this level value u 30 minutes or longer. cannot be accurately determined during AND Cold Shutdown or Refueling modes, b. ANY indication from the therefore this EAL was not included.
Containment Challenge Table (see below). (2) a. (Reactor vessel/RCS
[PWR] (1) a. Reactor vessel/RCS level Difference Global Comment #4 None or RPV [BWR]) level cannot cannot be monitored for be mo"nitored for 30 30 minutes or longer. minutes or longer. 35 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# AND AND Difference Global Comment #8, 9 & 13 V8 &V16 b. Core uncovery is indicated
: b. Core uncovery is indicated by ANY of the following:
by ANY of the following:
* (Site-specific radiation
* Containment High monitor) reading greater Radiation Monitor {1{2)-than (site-specific value) RE-126, RE-127, or RE-* Erratic source range 128) reading greater monitor indication than 100 R/hr [PWR]
* Erratic source range
* UNPLANNED increase in monitor indication (site-specific sump
* UNPLANNED increase in and/or tank) levels of Containment Sump A sufficient magnitude to OR Waste Holdup Tank indicate core uncovery levels of sufficient AND magnitude to indicate core uncovery C. ANY indication from the AND Containment Challenge C. ANY indication from Difference Global Comment #8 & 9 None Table (see below). Containment Challenge Table C-1 Containment Challenge Table ONTAINMENT CLOSURE not established*
Containment Challenge Table C Difference Global Comment #8 & 9 V17 xplosive mixture) exists inside containment
* CONTAINMENT CLOSURE not established*
NPLANNED increase in containment pressure
* 6% H2 concentration exists inside containn "Condary containment radiation monitor reading . UNPLANNED increase in containment pres ite specific value) [BWR]
* If CONTAINMENT CLOSURE is *If CONTAINMENT CLOSURE is Verbatim re-established prior to exceeding re-established prior to exceeding the 30-minute time limit, then the 30-minute time limit, then declaration of a General declaration of a General Intent and meaning of the EALs are not Emergency is not required.
Emergency is not required.
altered. 36 PBNP DEVIATIONS AND DIFFERENCES MATRIX INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS 37 PBNP DEVIATIONS AND DIFFERENCES MATRIX ,* Section NEI 99-01 Rev. 6 PBNP Change
* Justification Validation
# Recognition Category:
E-HUl EUl Difference PBNP changed Recognition Category to None maintain continuity with previous practice at PBNP. Global Comment #14 Initiating Condition:
Damage to a Damage to a loaded cask Verbatim None loaded cask CONFINEMENT CONFINEMENT BOUNDARY.
BOUNDARY.
Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) Damage to a loaded cask (1) Damage to a loaded cask Difference Global Comment #8, 9 & 12 V18 CONFINEMENT CONFINEMENT BOUNDARY as BOUNDARY as indicated by indicated by an on-contact an on-contact radiation radiation reading greater reading greater than (2 than the values shown below .... times the site-specific cask on the surface of the spent ::::> specific technical fuel cask. :I: I specification allowable w radiation level) on the 32 PT DSC surface of the spent fuel Front Surface 1700 mrem/ hr Door Centerline 400 mrem/ hr cask. End Shield Wall 12 mrem/ hr Exterior VSC-24 Sides 200 mrem/ hr Top 400 mrem/ hr Air Inlets 700 mrem/ hr Air Outlets 200 mrem/ hr Intent and meaning of the EALs are not altered. 38 PBNP DEVIATIONS AND DIFFERENCES MATRIX FISSION PRODUCT BARRIER ICS/EALS The following section is configured in a manner that is different from the Fission Product Barrier Tables in the PBNP EAL Technical Bases Document.
Where the Technical Bases Document evaluates all three fission product barriers simultaneously for a specific sub-category, this matrix presents each fission product barrier individually for all sub-categories.
The significance of this presentation is that where the fission product barrier table in the Technical Bases Document moves vertically through the categories, this matrix moves horizontally.
39 PBNP DEVIATIONS AND DIFFERENCES MATRIX Fission Product Barrier Emergency Classifications NEI 99-01 Rev. 6 PBNP Change Justification Validation#
Table 9-F-1: Recognition Category "F" Initiating Condition Matrix Alert Site Area General Emergency Emergency Any Loss or Loss or Loss of any two any Potential Potential barriers and Loss of either Loss of any Loss or Deleted per developer note. Mode the Fuel Clad two barriers.
Potential Loss Deleted Difference applicability carried over onto Table 9-None or RCS barrier. of the third F-1 EAL listing. barrier. Global Comment #10 Op Modes: Op Modes: Op Modes: Power Power Power Operation, Hot Operation, Operation, Hot Standby, Hot Standby, Standby, Startup, Hot Startup, Hot Startup, Hot Shutdown Shutdown Shutdown Table 9-F-2: BWR EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Deleted Difference Global Comment #4 None Barriers Table 9-F-3: PWR EAL Fission Product Barrier Table 9-F-1: EAL Fission Product Renumbered and re-labeled due to deletion ofTables 9-F-1 & 2. Table Thresholds for LOSS or POTENTIAL LOSS of Barrier Table Thresholds for LOSS or Difference Added None Barriers POTENTIAL LOSS of Barriers Global Comment #9 Basis Information For PWR EAL Fission Product Deleted Developer Notes Difference Transform generic NEI 99-01 guidance Barrier Table 9-F-3 Developer Notes. into PBNP specific application.
None Figure 9-F-4: PWR Containment Integrity or Figure 9-F-2: Containment Integrity or Renumbered and re-labeled due to Difference deletion of prior Tables. None Bypass Example Bypass Example Global Comment #9 40 PBNP DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Fuel Clad Barrier .. T~ble 9lF-1 NEI 99-01 Rev. 6 PBNP Change Justification Sub~Category Loss Potential Loss Loss Potential Loss Critica I *Safety Not Used Not Used A. Conditions requiring A. Conditions requiring Difference V19 -CSF status Functi~n Status entry into Core entry into Core consolidated into a Cooling RED Path Cooling ORANGE separate sub-category to ' {CSP C.1) are met. Path {CSP C.2) are : allow operators to quickly met. assess barrier status via i OR CSF status at one time. ' .. B . Conditions requiring " entry into Heat Sink V20 -CSF status used consistent with : ,., RED Path (CSP H.1) are met. recommendations for ,,,.' .. Westinghouse ERG plants. i. RCS or SG : Not Applicable A. RCS/reactor vessel Not Applicable Not Applicable Difference CSF status used consistent . Tube ~eakage level less than (site-with recommendations for ., specific level). Westinghouse ERG plants, *' ,' ,* but listed in separate sub-category 1. 2. Inadequate A. Core exit A. Core exit Sub-category not used Sub-category not used Difference CSF status consolidated Heat Removal thermocouple thermocouple into a separate sub-readings greater than readings greater category to allow (site-specific than (site-specific operators to quickly assess temperature value). temperature value). barrier status via CSF OR status at one time. B. Inadequate RCS heat removal CSF status used consistent capability via steam with recommendations for .. generators as Westinghouse ERG plants. : : indicated by (site-specific indications).
41 PBNP DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Fuel Clad Barrier : Table 9-F-1 NEI 99-01 Rev. 6 PBNP Change Justification
' Sub,~C~tegory Loss Potential Loss Loss Potential Loss t RCS Activity/
A. Containment Not Applicable A. Containment Not Applicable Difference V21 -Global Comment #9 i ~ontainment radiation monitor radiation monitor , Radiation: . reading greater than reading greater than ; (site-specific value). 577 R/hr indicated , .. ; '* OR on ANY of the '. ; . ' B. (Site-specific following . indications that
* 1{2)-RE-126
' '' reactor coolant ,.
* 1(2)-RE-127 activity is greater
* 1(2)-RE-128
.. than 300 &#xb5;Ci/gm dose ' ' ''i'. OR equivalent 1-131). B. 1{2)-RE-109 greater V22 -Global Comment #9 ;',, :.: than 4,500 mR/hr ,,, 4. Cpntainment Not Applicable Not Applicable Not Applicable Not Applicable Verbatim None . lntegritv.
or Bypass S.Other A. (site-specific as A. (site-specific as Sub-category not used Sub-category not used Difference Global Comment #9
* Indications applicable) applicable)
No other site-specific thresholds were identified for PBNP . . 6: Ertiergericy A. ANY condition in the A. ANY condition in the A. ANY condition in the A. ANY condition in the Verbatim None ' : Director .
* opinion of the opinion of the opinion of the opinion of the ' 1 Judgment ', Emergency Director Emergency Director Emergency Director Emergency Director ,. that indicates Loss of that indicates that indicates Loss of that indicates the Fuel Clad Barrier. Potentia I Loss of the the Fuel Clad Barrier. Potential Loss of the ,, ' ' Fuel Clad Barrier. Fuel Clad Barrier. : " 42 _J PBNP DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of RCS Barrier Table9-F-1 NEI 99-01 Rev. 6 PBNP Change Justification
* Sub::category Loss Potential Loss Loss Potential Loss Critical Safety Not Used Not Used Not Applicable A. Conditions requiring Difference V20 -CSF status Function Status entry into Heat Sink consolidated into a RED Path (CSP H.1) separate sub-category to : are met. allow operators to quickly OR assess barrier status via CSF status at one time. . , B . Conditions requiring entry into RCS V23 -CSF status used : Integrity RED Path consistent with .. (CSP P.1) are met . recommendations for Westinghouse ERG plants. 1. RCS orSG* A. An automatic or A. Operation of a A. An automatic or A. Operation of a Difference Global Comment #9 *Tube manual ECCS (SI) standby charging manual ECCS (SI) standby charging .Leakage actuation is required (makeup) pump is actuation is required (makeup) pump is CSF status consolidated by EITHER of the required by EITHER by EITHER of the required by EITHER into a separate sub-following:
of the following:
following:
of the following:
category to allow . . 1 . UNISOLABLE RCS 1. UNISOLABLE
: 1. UNISOLABLE RCS 1. UNISOLABLE operators to quickly assess leakage RCS leakage leakage RCS leakage barrier status via CSF OR OR OR OR status at one time. 2. SG tube RUPTURE. 2. SG tube 2. SG tube 2. SG tube leakage. leakage. RUPTURE. CSF status used consistent OR with recommendations for B. RCS cooldown rate Westinghouse ERG plants. greater than (site-specific pressurized thermal shock ' criteria/limits defined by site-specific indications).
43 PBNP DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of RCS Barrier Table,9-F-1 NEI 99-01 Rev. 6
* PBNP Change Justification
'*1,1' :* l* *\Inadequate Not Applicable A. Inadequate RCS Sub-category not used Sub-category not used Difference CSF status consolidated
.. ":heat Removal heat removal '' into a separate sub-'' *' '' capability via steam category to allow generators as operators to quickly ,. indicated by (site-assess barrier status via ,, '' '' ': specific CSF status at one time. .. ,, indications).
: 3. 'RCS Activity/
A. Containment Not Applicable Containment radiation Not Applicable Difference V21 -Global Comment
* Containment radiation monitor monitor reading #9 . *
* Radiation . " reading greater than greater than 11 R/hr on ,, (site-specific value). ANY of the following:
'' ',
* 1{2)-RE-126
:
* 1(2)-RE-127
* 1(2)-RE-128
,, 4. Containment Not Applicable Not Applicable Not Applicable Not Applicable Verbatim None . Integrity or . Bypass 5. Other A. (site-specific as A. (site-specific as Sub-category not used Sub-category not used Difference Global Comment #9 ' Indications applicable) applicable)
No other site-specific
'' " thresholds were ,' '' ''' identified for PBNP. " 6 *. Emergency A. ANY condition in the A. ANY condition in A. ANY condition in A. ANY condition in Verbatim None 'Director opinion of the the opinion of the the opinion of the the opinion of the ':,Judgment Emergency Director Emergency Director Emergency Director Emergency
.,i that indicates Loss of that indicates that indicates Loss Director that : ; .,,* *'" Potential Loss of of the RCS Barrier. indicates Potential . ' the RCS Barrier . the RCS Barrier. Loss of the RCS " ' ' Barrier. 44 PBNP DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Containment Barrier 'Table 9-f-2' NEI 99-01 Rev. 6 PBI\JP Change Justification Sub~Category Loss Potential Loss Loss Potential Loss * ~ritica,I Safety Not Used Not Used Not Applicable A. 1. Conditions Difference V19 -Global comment #9 Function Status requiring entry CSF status consolidated into Core into a separate sub-' ,. Cooling RED category to allow ' ... Path (CSP C.1) operators to quickly assess ' : ' are met. barrier status via CSF *, " status at one time. AND ' *' 2. CSP C.1 not ,*, effective within CSF status used consistent 15 minutes. with recommendations for Westinghouse ERG plants. 1. R(;S or SG A. A leaking or Not Applicable A. A leaking or Not Applicable Verbatim None Tube Leakage RUPTURED SG is RUPTURED SG is FAULTED outside of FAULTED outside of containment . containment.
.. 2: Inadequate , Not Applicable A 1 (Site-specific criteria Sub-category not used Sub-category not used Difference CSF status consolidated heat Removal for entry into core into a separate sub-cooling restoration category to allow procedure) operators to quickly assess AND barrier status via CSF 2 Restoration status at one time. *, : procedure not effective within 15 . . ,, minutes . ,, ,,: 45 PBNP DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Containment Barrier , : Table 9-F,-2 NEI 99-01 Rev. 6 PBNP Change Justification
\su6~c:ategory.;:
Loss Potential Loss Loss Potential Loss 3 .. RCS .Activity/
Not Applicable A. Containment Not Applicable A. Containment Difference V21-Global Comment #9 : : ' Containment radiation monitor radiation monitor
* Radiation reading greater than reading greater than (site-specific value). 18,500 R/hr ' indicated on ANY of ,, the following.
* 1(2)-RE-126
* 1{2)-RE-127
* 1(2)-RE-128
: 4. Containment A. Containment isolation A. Containment A. Containment A. Containment Difference V24 -Global Comment #9 . Integrity or is required AND ' pressure greater isolation is required pressure greater '. Bypass EITHER of the than (site-specific AND EITHER of the than 60 psig. following:
value) following:
OR *, OR V17 '' 1. Containment B. 6% H 2 inside 1-, ,: _, 1. Containment B. Explosive mixture ' ' > exists inside integrity has containment.
'' " integrity has been been lost based OR lost based on containment on Emergency Emergency OR Director C. 1. Containment V25 ,,,.* Director C 1.Containment pressure greater '' judgment.
'' judgment.
pressure greater than 25 psig. OR than (site-specific OR AND 2. UNISOLABLE pressure setpoint)
: 2. UNISOLABLE
: 2. Less than one full pathway from the AND pathway from train of I containment to 2. Less than one full the containment
' '* : depressurization
*: '' ,, ', the environment train of (site-specific to the ,, ' equipment is ,, 1, environment
: exists. system or operating per " OR equipment) is exists. design for 15 B. Indications of RCS operating per OR minutes or leakage outside of design for 15 B. Indications of RCS longer. containment minutes or longer. leakage outside of containment.
46 PBNP DEVIATIONS AND DIFFERENCES MATRIX Thresholds for LOSS or POTENTIAL LOSS of Containment Barrier Table 9-F-2 NEI 99-01 Rev. 6 PBNP Change Justific_ation Sub-Category Loss Potential Loss Loss Potential Loss 5.0ther A. (site-specific as A. (site-specific as Sub-category not used Sub-category not used Difference Global Comment #9 Indications applicable) applicable)
No other site-specific thresholds were identified for PBNP. 6.Emergency A. ANY condition in the B. ANY condition in the C. ANY condition in the D. ANY condition in the Verbatim None Director opinion of the opinion of the opinion of the opinion of the Judgment Emergency Director Emergency Director Emergency Director Emergency Director that indicates Loss of that indicates that indicates Loss that indicates the Containment Potentia I Loss of the of the Containment Potential Loss of the Barrier. Containment Barrier. Containment Barrier. Barrier. 47 PBNP DEVIATIONS AND DIFFERENCES MATRIX HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 48 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
HU1 HU1 Verbatim Global Comment #14 None Initiating Condition:
Confirmed Confirmed SECURITY CONDITION Verbatim None SECURITY CONDITION or threat. or threat. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) A SECURITY CONDITION (1) A SECURITY CONDITION that Difference Global Comment #9 & 12 None that does not involve a does not involve a HOSTILE HOSTILE ACTION as ACTION as reported by the reported by the (site-Security Shift Supervisor.
specific security shift supervision) . .... ::::, (2) Notification of a credible (2) Notification of a credible Difference Global Comment #8 & 9 None :c security threat directed at security threat directed at the site. PBNP. (3) A validated notification (3) A validated notification from Difference Global Comment #8 & 9 None from the NRC providing the NRC providing information of an aircraft information of an aircraft threat. threat. Intent and meaning of the EALs are not altered. 49 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
HU2 HU2 Verbatim None Initiating Condition:
Seismic Seismic event greater than OBE Verbatim None event greater than OBE levels. levels. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) Seismic event greater than (1) Seismic event greater than Difference Global Comment #9 & 12 V26 Operating Basis Operating Basis Earthquake Earthquake
{OBE) as (OBE) as indicated by seismic Sentence structure altered from list to indicated by: monitor indication of ground accommodate a single indication as defined (site-specific indication that a acceleration greater than: by PBNP procedure NP 7.2.29. seismic event met or exceeded
* 0.06 g horizontal N :::::> OBE limits) OR :c
* 0.04 g vertical Intent and meaning of the EALs are not altered. 50 PBNP DEVIATIONS AND DIFFERENCES MATRIX , Section NEI 99-0:L Rev. 6 PBNP Change Justification Validation
# Recognition Category:
HU3 HU3 Verbatim Global Comment #14 None Initiating Condition:
Hazardous Hazardous event. Verbatim None event. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) A tornado strike within the (1) A tornado strike within the Verbatim Global Comment #12 None PROTECTED AREA. PROTECTED AREA. (2) Internal room or area (2) Internal room or area Verbatim None flooding of a magnitude flooding of a magnitude sufficient to require sufficient to require manual manual or automatic or automatic electrical electrical isolation of a isolation of a SAFETY SYSTEM SAFETY SYSTEM component needed for the component needed for the current operating mode. current operating mode. (3) Movement of personnel (3) Movement of personnel Verbatim None within the PROTECTED within the PROTECTED AREA AREA is impeded due to an is impeded due to an offsite M offsite event involving event involving hazardous
::::, hazardous materials (e.g., materials (e.g., an offsite :c an offsite chemical spill or chemical spill or toxic gas toxic gas release).
release).
(4) A hazardous event that (4) A hazardous event that Verbatim None results in on-site results in on-site conditions conditions sufficient to sufficient to prohibit the plant prohibit the plant staff staff from accessing the site from accessing the site via via personal vehicles.
personal vehicles.
(5) (Site-specific list of natural (5) Lake level greater than or Difference Global Comment #9 V14 or technological hazard equal to +8.0 ft. events) (Plant elevation)
(6) Pump bay level less than -Difference Global Comment #9 V14 15.0 ft. Added as a 6th EAL versus a list in EAL #5 to maintain consistent format Intent and meaning of the EALs are not altered. 51 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category: HU4 HU4 Verbatim Global Comment #14 None Initiating Condition:
FIRE FIRE potentially degrading the Verbatim None potentially degrading the level of level of safety of the plant. safety of the plant. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) a. A FIRE is NOT extinguished (1) a. A FIRE is NOT extinguished Difference Global Comment #12 & 13 None within 15-minutes of ANY within 15-minutes of ANY of the following FIRE of the following FIRE detection indications:
detection indications:
* Report from the field
* Report from the field (i.e., visual observation) (i.e., visual
* Receipt of multiple observation) (more than 1) fire
* Receipt of multiple alarms or indications (more than 1) fire
* Field verification of a alarms or indications single fire alarm
* Field verification of a AND single fire alarm AND b. The FIRE is located within b. The FIRE is located within Difference Global Comment #8, 9, & 13 None ANY of the following plant ANY Table H-1 plant rooms or areas: rooms or areas: (site-specific list of plant rooms <:t or areas) ::, :I: (2) a. Receipt of a single fire (2) a. Receipt of a single fire Deviation PBNP proposes to make EAL HU4.2 V35 alarm (i.e., no other alarm with no other applicable to a single fire alarm in indications of a FIRE). indications of a FIRE. Containment during operation in Modes 3 AND AND and 4 only. Note added to IC that a b. The FIRE is located within b. The FIRE is located within "Containment fire alarm is considered valid ANY of the following plant ANY Table H-1 plant rooms upon receipt of multiple zones (more than rooms or areas: or areas except 1) on the FACP system (this note is (site-specific list of plant rooms Containment in Modes 1 applicable in Modes 1 and 2 only)." or areas) and 2 (see Note above): Global Comment #8, 9 & 13 AND AND 52 
. Secticm NEI 99-01 Rev. 6 c. The existence of a FIRE is not verified within 30-minutes of alarm receipt. (3) A FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area] PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.
(4) A FIRE within the plant or ISFSI [for plants with an ISFSI outside the plant Protected Area] PROTECTED AREA that requires firefighting support by an offsite fire re?ponse agency to extinguish.
PBNP DEVIATIONS AND DIFFERENCES MATRIX PBNP Change Justification
: c. The existence of a FIRE is not verified within 30-minutes of alarm receipt. (2) A FIRE within the plant or Difference Global Comment #9 ISFSI PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication (3) A FIRE within the plant or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
Table H-1 Areas Control Room Containment PAB GOS building 13.BkV Building Cable Spreading Room Vital Switchgear Room AFW Pump Room G-01/02 Rooms EDG Building Service Water Pump Rooms Fa~ade 85' 53 Difference Global Comment #9 Difference Global Comment #8 & 9 Basis revised to include clarification of Containment fire alarms, and to include NFPA-805 in the discussion of Appendix R basis for the EAL thresholds.
Intent and meaning of the EALs are not altered. Validation
# None V27 None PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
HU7 HU7 Verbatim Global Comment #14 None Initiating Condition:
Other Other conditions exist which in Verbatim None conditions exist which in the the judgment of the Emergency judgment of the Emergency Director warrant declaration of a Director warrant declaration of a (NO) UE. (NO) UE. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) Other conditions exist (1) Other conditions exist Verbatim Global Comment #12 None which in the judgment of which in the judgment of the Emergency Director the Emergency Director indicate that events are in indicate that events are in ..... progress or have occurred progress or have occurred :) which indicate a potential which indicate a potential
:::c degradation of the level of degradation of the level of safety of the plant or safety of the plant or indicate a security threat indicate a security threat to facility protection has to facility protection has been initiated.
No releases been initiated.
No releases of radioactive material of radioactive material requiring offsite response requiring offsite response or monitoring are or monitoring are expected unless further expected unless further degradation of safety degradation of safety systems occurs. systems occurs. 54 PBNP DEVIATIONS AND DIFFERENCES MATRIX Sectiqn NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
HAl HAl Verbatim Global Comment #14 None Initiating Condition:
HOSTILE HOSTILE ACTION within the Verbatim None ACTION within the OWNER OWNER CONTROLLED AREA or CONTROLLED AREA or airborne airborne attack threat within 30 attack threat within 30 minutes. minutes. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) A HOSTILE ACTION is (1) A HOSTILE ACTION is Difference Global Comment #9 & 12 None occurring or has occurred occurring or has occurred within the OWNER within the OWNER ..-1 <C CONTROLLED AREA as CONTROLLED AREA as :::c: reported by the (site-reported by the Security specific security shift Shift Supervisor.
supervision).
(2) A validated notification (2) A validated notification Verbatim from NRC of an aircraft from NRC of an aircraft attack threat within 30 attack threat within 30 minutes of the site. minutes of the site. Intent and meaning of the EALs are not altered. 55 
:
* Section r..n c::i: :::c NEI 99-01 Rev. 6 Recognition Category:
HAS Initiating Condition:
Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability:
All (1) a. Release of a toxic, corrosive, asphyxiant or flammable gas into any of the following plant rooms or areas: (site-specific list of plant rooms or areas with entry-related mode applicability identified)
AND b. Entry into the room or area is prohibited or impeded. PBNP DEVIATIONS AND DIFFERENCES MATRIX PBNP Change HAS Verbatim Gaseous release impeding access Verbatim to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability:
All Verbatim {1) a. Release of a toxic, Difference corrosive, asphyxiant or flammable gas into any of the following plant rooms or areas: Table H-2 SAFE OPS, S/D, C/D AREAS .. _}\rea/Building Ul VCT Area U2 VCT Area Ul Primary Sample area U2 Primary Sample area CCW HX Room C-59 area Pipeway 2, 8 ft. Elev. Pipeway 3, 8 ft. Elev. 1/2B32 MCC Area AND . MODE .. 3/4/5 3/4/ 5 3 3 4/5 3/4/5 3/4 3/4 4 a. Entry into the room or area is prohibited or impeded. 56 Verbatim Justification Global Comment #8, 9 & 12 Intent and meaning of the EALs are not altered. Validation
# None None None V28 None PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
HA6 HA6 Verbatim None Initiating Condition:
Control Control Room evacuation Verbatim None Room evacuation resulting in resulting in transfer of plant transfer of plant control to control to alternate locations.
alternate locations.
Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None u:, (1) An event has resulted in (1) An event has resulted in plant Difference Global Comment #9 & 12 None c::( :::c: plant control being control being transferred transferred from the from the Control Room to Control Room to (site-AOP local control stations.
specific remote shutdown panels and local control stations).
Intent and meaning of the EALs are not altered. 57 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category: HA7 HA7 Verbatim Global Comment #14 None Initiating Condition:
Other Other conditions exist which in Verbatim None conditions exist which in the the judgment of the Emergency judgment of the Emergency Director warrant declaration of Director warrant declaration of an Alert. an Alert. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) Other conditions exist (1) Other conditions exist which, Verbatim Global Comment #12 None which, in the judgment of in the judgment of the the Emergency Director, Emergency Director, indicate indicate that events are in that events are in progress or progress or have occurred have occurred which involve which involve an actual or an actual or potential " potential substantial substantial degradation of c::( :::c degradation of the level of the level of safety of the safety of the plant or a plant or a security event that security event that involves probable life involves probable life threatening risk to site threatening risk to site personnel or damage to site Intent and meaning of the EALs are not personnel or damage to equipment because of altered site equipment because of HOSTILE ACTION. Any HOSTILE ACTION. Any releases are expected to be releases are expected to limited to small fractions of be limited to small the EPA Protective Action fractions of the EPA Guideline exposure levels. Protective Action Guideline exposure levels. 58 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
HS1 HS1 Verbatim Global Comment #14 None Initiating Condition:
HOSTILE HOSTILE ACTION within the Verbatim None ACTION within the PROTECTED PROTECTED AREA. AREA. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) A HOSTILE ACTION is (1) A HOSTILE ACTION is Difference Global Comment #9 & 12 None ,-1 occurring or has occurred occurring or has occurred II) within the PROTECTED within the PROTECTED AREA :::c AREA as reported by the as reported by the Security (site-specific security shift Shift Supervisor.
supervision).
Intent and meaning of the EALs are not altered. 59 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
HS6 HS6 Verbatim None Initiating Condition:
Inability to Inability to control a key safety Verbatim None control a key safety function function from outside the Control from outside the Control Room. Room. Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None Note: The Emergency Director Note: The Emergency Director Global Comment #9 None should declare the Site Area should declare the Site Area Emergency promptly upon Emergency promptly upon 15 minutes used per Developers Note determining that (site specific determining that 15 minutes has number of) minutes has been been exceeded, or will likely be exceeded, or will likely be exceeded.
exceeded.
(1) a. An event has resulted in (1) a. An event has resulted in Difference Global Comment #9 & 12 None plant control being plant control being transferred from the transferred from the U) Control Room to (site-Control Room to AOP local V'l :::c specific remote shutdown control stations.
panels and local control stations).
AND AND Difference Global Comment #9 None b. Control of ANY of the b. Control of ANY of the following key safety following key safety functions is not functions is not reestablished within (site-reestablished within 15 specific number of minutes. minutes).
* Reactivity Control
* Reactivity control
* Core Cooling
* Core cooling [PWR] /
* RCS Heat Removal RPV water level [BWR]
* RCS heat removal Intent and meaning of the EALs are not altered. 60 PBNP DEVIATIONS AND DIFFERENCES MATRIX . SectiQn NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
HS7 Recognition Category:
HS7 Verbatim Global Comment #14 None Initiating Condition:
Other Initiating Condition:
Other Verbatim None conditions exist which in the conditions exist which in the judgment of the Emergency judgment of the Emergency Director warrant declaration of a Director warrant declaration of a Site Area Emergency.
Site Area Emergency.
Operating Mode Applicability:
All Operating Mode Applicability:
Verbatim None ALL (1) Other conditions exist (1) Other conditions exist which Verbatim Global Comment #12 None which in the judgment of in the judgment of the the Emergency Director Emergency Director indicate indicate that events are in that events are in progress or progress or have occurred have occurred which involve which involve actual or actual or likely major failures likely major failures of of plant functions needed for plant functions needed for protection of the public or " V) protection of the public or HOSTILE ACTION that results :c HOSTILE ACTION that in intentional damage or results in intentional malicious acts, (1) toward site damage or malicious acts, personnel or equipment that (1) toward site personnel could lead to the likely failure or equipment that could of or, (2) that prevent lead to the likely failure of effective access to equipment or, (2) that prevent needed for the protection of Intent and meaning of the EALs are not effective access to the public. Any releases are altered equipment needed for the not expected to result in protection of the public. exposure levels which exceed Any releases are not EPA Protective Action expected to result in Guideline exposure levels exposure levels which beyond the site boundary.
exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
61 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
HG1 HG1 Verbatim Global Comment #14 None Initiating Condition:
HOSTILE HOSTILE ACTION resulting in loss Verbatim None ACTION resulting in loss of of physical control of the facility.
physical control of the facility.
Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) a. A HOSTILE ACTION is (1) a. A HOSTILE ACTION is Difference Global Comment #9 & 12 None occurring or has occurred occurring or has occurred within the PROTECTED within the PROTECTED AREA as reported by the AREA as reported by the (site-specific security shift Security Shift Supervisor.
supervision).
AND AND Difference Global Comment #9 None b. EITHER of the following
: b. EITHER of the following
.-1 has occurred:
has occurred: ::i:: 1. ANY of the following
: 1. ANY of the following safety functions cannot safety functions cannot be controlled or be controlled or maintained.
maintained.
* Reactivity control
* Reactivity Control
* Core cooling [PWR] /
* Core Cooling RPV water level
* RCS Heat Removal [BWR]
* RCS heat removal OR OR Verbatim None 2. Damage to spent fuel 2. Damage to spent fuel has occurred or is has occurred or is IMMINENT.
IMMINENT.
Intent and meaning of the EALs are not altered. 62 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category: HG7 HG7 Verbatim Global Comment #14 None Initiating Condition:
Other Other conditions exist which in Verbatim None conditions exist which in the the judgment of the Emergency judgment of the Emergency Director warrant declaration of a Director warrant declaration of a General Emergency.
General Emergency.
Operating Mode Applicability:
All Operating Mode Applicability:
All Verbatim None (1) Other conditions exist (1) Other conditions exist which Verbatim Global Comment #12 None which in the judgment of in the judgment of the the Emergency Director Emergency Director indicate indicate that events are in that events are in progress or progress or have occurred have occurred which involve which involve actual or actual or IMMINENT " IMMINENT substantial substantial core degradation core degradation or or melting with potential for :c melting with potential for loss of containment integrity loss of containment or HOSTILE ACTION that integrity or HOSTILE results in an actual loss of ACTION that results in an physical control of the actual loss of physical facility.
Releases can be control of the facility.
reasonably expected to Releases can be reasonably exceed EPA Protective Action expected to exceed EPA Guideline exposure levels Protective Action Guideline offsite for more than the exposure levels offsite for immediate site area. more than the immediate site area. Intent and meaning of the EALs are not altered 63 PBNP DEVIATIONS AND DIFFERENCES MATRIX SYSTEM MALFUNCTION ICS/EALS 64 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
SUl SUl Verbatim None Initiating Condition:
Loss of all Loss of all offsite AC power Verbatim None offsite AC power capability to capability to emergency buses for emergency buses for 15 minutes 15 minutes or longer. or longer. Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None .-l 2,3,4 :) Power Operation, Startup, Hot c./) Standby, Hot Shutdown (1) Loss of ALL offsite AC (1) Loss of ALL offsite AC power Difference Global Comment #9 & 12 None power capability to (site-capability to 1{2)-A05 and specific emergency buses) 1(2)-A06 for 15 minutes or for 15 minutes or longer. longer. Intent and meaning of the EALs are not altered. 65 PBNP DEVIATIONS AND DIFFERENCES MATRIX
* Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
SU2 SU3 Verbatim Global Comment #14, Renumbered to align None with other similar ICs Initiating Condition:
UNPLANNED UNPLANNED loss of Control Verbatim None loss of Control Room indications Room indications for 15 minutes for 15 minutes or longer. or longer. Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3,4 Standby, Hot Shutdown (1) a. An UNPLANNED event (1) a. An UNPLANNED event Difference Global Comment #9 & 12 None results in the inability to results in the inability to monitor one or more of monitor one or more of the following parameters the following parameters from within the Control from within the Control Room for 15 minutes or Room for 15 minutes or longer. longer. Difference Global Comment #9 None N [BWR parameter
[PWR
* Reactor Power ::::> list] para meter list] V)
* RCS/ Pressurizer Reactor Power Reactor Power Level RPV Water Level RCS Level RPV Pressure RCS Pressure
* RCS/ Pressurizer Primary In-Core/Core Pressure Containment Exit
* Core Exit Pressure Temperature Temperature Suppression Pool Levels in at least Level (site-specific
* Level in at least one number) two steam generator steam
* Steam Generator generators Auxiliary Feed Water Suppression Pool Steam Flow Temperature Generator Auxiliary or Emergency Feed Water Flow Intent and meaning ofthe EALs are not altered. 66 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
SU3 SU4 Verbatim Renumbered to align with other similar ICs None Initiating Condition:
Reactor Reactor coolant activity greater Verbatim None coolant activity greater than than Technical Specification Technical Specification allowable allowable limits. limits. Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3,4 Standby, Hot Shutdown (1) (Site-specific radiation (1) Failed Fuel Monitor 1(2)-RE-Difference Global Comment #9 & 12 V30 monitor) reading greater 109 reading greater than 750 than (site-specific value). mR/hr (2) Sample analysis indicates (2) Sample analysis indicates Difference Global Comment #9 V31 that a reactor coolant that a RCS Specific Activity activity value is greater value is greater than an than an allowable limit allowable limit specified in m specified in Technical Technical Specifications as => 1/) Specifications.
indicated by ANY of the following conditions:
: a. Dose Equivalent 1-131 greater than 50 &#xb5;Ci/gm OR b. Dose Equivalent 1-131 greater than 0.5 &#xb5;Ci/gm but less than or equal to 50 &#xb5;Ci/gm for greater than 48 hours OR C. Dose Equivalent Xe-133 greater than 300 &#xb5;Ci/gm for greater than 48 hours Intent and meaning of the EALs are not altered. 67 PBNP DEVIATIONS AND DIFFERENCES MATRIX
* Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
SU4 SUS Verbatim Global Comment #14, Renumbered to align None with other similar ICs Initiating Condition:
RCS leakage RCS leakage for 15 minutes or Verbatim None for 15 minutes or longer. longer. Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3, 4 Standby, Hot Shutdown (1) RCS unidentified or (1) RCS unidentified or pressure Difference Global Comment #9 & 12 None pressure boundary leakage boundary leakage greater greater than (site-specific than 10 gpm for 15 minutes value) for 15 minutes or or longer. longer. <::I' (2) RCS identified leakage (2) RCS identified leakage greater Difference Global Comment #9 V32 ::::i V) greater than (site-specific than 25 gpm for 15 minutes value) for 15 minutes or or longer. longer. (3) Leakage from the RCS to a (3) Leakage from the RCS to a Difference Global Comment #9 .None location outside location outside containment, Added specific language about SG tube containment greater than or Steam Generator tube leakage to the EAL at the request of 25 gpm for 15 minutes or leakage, greater than 25 gpm operators longer. for 15 minutes or longer. Intent and meaning of the EALs are not altered. 68 
-------------------
PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
SUS SU6 Verbatim Global Comment #14, Renumbered to align None with other similar !Cs Initiating Condition:
Automatic or Automatic or manual trip fails to Difference Global Comment #4 None manual (trip [PWR] / scram shutdown the reactor. [BWR]) fails to shutdown the reactor. Operating Mode Applicability:
Operating Mode Applicability:
1 Difference Global Comment #10 None Power Operation (1) a. An automatic (trip [PWR] / (1) a. An automatic trip did not Difference Global Comment #4 & 12 None scram [BWR]) did not shutdown the reactor. shutdown the reactor. AND AND Verbatim None b. A subsequent manual b. A subsequent manual action taken at the reactor action taken at the reactor control consoles is control consoles is successful in shutting successful in shutting down the reactor. down the reactor. in (2) a. A manual trip ([PWR] / (2) a. A manual trip did not Difference Global Comment #4 None ::, V) scram [BWR]) did not shutdown the reactor. shutdown the reactor. AND AND Verbatim None b. EITHER of the following:
: b. EITHER of the following:
: 1. A subsequent manual 1. A subsequent manual action taken at the action taken at the reactor control consoles reactor control consoles is successful in shutting is successful in shutting down the reactor. down the reactor. OR OR Difference Global Comment #4 None 2. A subsequent automatic
: 2. A subsequent automatic Added Note to basis to define PBNP "at the (trip [PWR] / scram trip is successful in reactor control consoles" as meaning the [BWR]) is successful in shutting down the reactor trip pushbuttons on the following shutting down the reactor. panels: reactor.
* Unit 1 on panels 1C04 and COl
* Unit 2 on panels 2C04 and CO2 Intent and meaning of the EALs are not altered. 69 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 t>BNP Change Justification*
Validation
# Recognition Category:
SU6 SU7 Verbatim Renumbered to align with other similar !Cs None Initiating Condition:
Loss of all Loss of all onsite or offsite Verbatim None onsite or offsite communications communications capabilities.
capabilities.
Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3,4 Standby, Hot Shutdown (1) Loss of ALL of the following (1) Loss of ALL of the following Difference Global Comment #9, 12 & 13 None Onsite communication Onsite communication methods: methods: (site-specific list of
* Plant Public Address System communications methods) (Gai-Tronics)
* Commercial Phones
* PBX Phones
* Security Radio
* Portable Radios I.C (2) Loss of ALL of the following (2) Loss of ALL of the following Difference Global Comment #9 & 13 None => ORO communications offsite response organization 1/) methods: communications methods: (site-specific list of
* Nuclear Accident Reporting communications methods) System (NARS)
* Commercial Phones
* PBX Phones
* Satellite Phones e Manitowoc County Sheriffs Department Radio (3) Loss of ALL of the following (2) Loss of ALL ofthe following Difference Global Comment #9 & 13 None NRC communications NRC communications methods: methods: (site-specific list of
* FTS Phone System communications methods)
* Commercial Phones
* PBX Phones
* Satellite Phones Intent and meaning of the EALs are not altered. 70 PBNP DEVIATIONS AND DIFFERENCES MATRIX
* Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
SU7 SU8 Verbatim Renumbered to align with other similar ICs None Initiating Condition:
Failure to Failure to isolate containment or Difference Global Comment #4 None isolate containment or loss of loss of containment pressure containment pressure control. control. [PWR] Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3,4 Standby, Hot Shutdown (1) a. Failure of containment to (1) a. Failure of containment to Verbatim Global Comment #12 None isolate when required by isolate when required by an actuation signal. an actuation signal. AND AND Verbatim None " b. ALL required penetrations
: b. ALL required penetrations
:::) VI are not closed within 15 are not closed within 15 minutes of the actuation minutes of the actuation signal. signal. (2) a. Containment pressure (2) a. Containment pressure Difference Global Comment #9 V25 greater than (site-specific greater than 25 psig. pressure).
AND AND Difference Global Comment #9 None b. Less than one full train of b. Less than one full train of (site-specific system or Containment Cooling equipment) is operating System equipment is per design for 15 minutes operating per design for 15 or longer. minutes or longer. Intent and meaning of the EALs are not altered. 71 PBNP DEVIATIONS AND DIFFERENCES MATRIX *. section NEI 99-01.Rev.
6 PBNP Change Justification Validation
# Recognition Category:
SAl SAl Verbatim Global Comment #14 None Initiating Condition:
Loss of all Loss of all but one AC power Verbatim None but one AC power source to source to emergency buses for 15 emergency buses for 15 minutes minutes or longer. or longer. Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3,4 Standby, Hot Shutdown (1) a. AC power capability to (1) a. AC power capability to Difference Global Comment #9, 12 & 13 None .... (site-specific emergency 1(2)-A05 AND 1(2)-A06 <( 1/'l buses) is reduced to a is reduced to a single single power source for 15 power source for 15 minutes or longer. minutes or longer. AND AND Difference Global Comment #13 None b. Any additional single b. Any additional single power source failure will power source failure will result in a loss of all AC result in a loss of ALL AC power to SAFETY SYSTEMS. power to SAFETY SYSTEMS. Intent and meaning of the EALs are not altered. 72 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification
* Validation
# Recognition Category: SA2 SA3 Verbatim Global Comment #14, Renumbered to align None with other similar I Cs Initiating Condition:
UNPLANNED UNPLANNED loss of Control Verbatim None loss of Control Room indications Room indications for 15 minutes for 15 minutes or longer with a or longer with a significant significant transient in progress.
transient in progress.
Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3,4 Standby, Hot Shutdown (1) a. An UNPLANNED event (1) a. An UNPLANNED event Verbatim Global Comment #12 None results in the inability to results in the inability to monitor one or more of monitor one or more of the following parameters the following parameters from within the Control from within the Control Room for 15 minutes or Room for 15 minutes or longer. longer.
* Reactor Power Difference Global Comment #4, 8 & 9 None [BWR [PWR parameter
* RCS/ Pressurizer N parameter list] list] Level c:( Reactor Power Reactor Power RCS/ Pressurizer V)
* RPV Water RCS Level Level Pressure RPV Pressure RCS Pressure
* Core Exit Primary In-Core/Core Exit Temperature Containment Temperature Levels in at least one
* Pressure Suppression Levels in at least steam generator Pool Level (site-specific
* Steam Generator number) steam Auxiliary Feed Water generators Flow Suppression Steam Generator Pool Auxiliary or Temperature Emergency Feed Water Flow 73 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section . NEI 99-01 Rev. 6 PBNP Change Justification ,Validation
# AND AND b. ANY of the following
: b. ANY of the following Difference Global Comment #4, 9 None transient events in transient events in progress.
progress.
* Automatic or manual
* Automatic or manual run back greater than run back greater than 25% thermal reactor 25% thermal reactor power power
* Electrical load rejection
* Electrical load greater than 25% full rejection greater than electrical load 25% full electrical
*. Reactor scram [BWR] / load trip [PWR]
* Reactor trip a ECCS (SI) actuation
* SI actuation
* Thermal power oscillations greater than 10% [BWR] Intent and meaning of the EALs are not altered. 74 PBNP DEVIATIONS AND DIFFERENCES MATRIX
* Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
SAS SA6 Verbatim Renumbered to align with other similar ICs None Initiating Condition:
Automatic or Automatic or manual trip fails to Difference Global Comment #4 & 9 None manual (trip [PWR] / scram shutdown the reactor, and [BWR]) fails to shutdown the subsequent manual actions taken reactor, and subsequent manual in the Control room are not actions taken at the reactor successful in shutting down the control consoles are not reactor. successful in shutting down the reactor. Operating Mode Applicability:
Operating Mode Applicability:
1 Difference Global Comment #10 None u, Power Operation
<C (1) a. An automatic or manual (1) a. An automatic or manual Difference Global Comment #4, 9 & 12 V) None (trip [PWR] / scram [BWR]) trip did not shutdown the did not shutdown the reactor. reactor. AND AND Verbatim None b. Manual actions taken at b. Manual actions taken at the reactor control the reactor control consoles are not successful consoles are not in shutting down the successful in shutting reactor. down the reactor. Added Note to basis to define PBNP "at the reactor control consoles" as meaning the reactor trip pushbuttons on the following panels:
* Unit 1 on panels 1C04 and COl
* Unit 2 on panels 2C04 and CO2 Intent and meaning of the EALs are not altered. 75 PBNP DEVIATIONS AND DIFFERENCES MATRIX
* Section
* NEI 99.-01 Rev .. 6 . PBNP Change Justification*
* .Validation#
Recognition Category:
SA9 SA9 Verbatim Global Comment #14 None Initiating Condition:
Hazardous Hazardous event affecting a Verbatim None event affecting a SAFETY SYSTEM SAFETY SYSTEM needed for the needed for the current operating current operating mode. mode. Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3,4 Standby, Hot Shutdown (1) a. The occurrence of ANY of (1) a. The occurrence of ANY of Difference Global Comment #12 & 13 None the following hazardous the following hazardous events: events:
* Seismic event
* Seismic event Difference Global Comment #8 & 9 V14 (earthquake) (earthquake)
* Internal or external
* Internal or flooding event external flooding
* High winds or tornado event "' strike
* High winds or <( II')
* FIRE tornado strike
* EXPLOSION
* FIRE * (site-specific hazards)
* EXPLOSION
* Other events with
* Lake level greater similar hazard than or equal to characteristics as +9.0 ft. (Plant determined by the Shift elevation)
Manager
* Pump bay level less than -19.0 ft.
* Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director 76 PBNP DEVIATIONS AND DIFFERENCES MATRIX Se<;tion NEI 99*-01 Rev. 6 PBNP Change Justification Validation
# AND AND b. EITHER of the following:
: b. 1. Event damage has Deviation Adopted the revised EAL structure and V15 1. Event damage has caused indications of wording provided in proposed EAL FAQ caused indications of degraded 2016-02. degraded performance performance in one in at least one train of a train of a SAFETY SAFETY SYSTEM needed SYSTEM needed for for the current the current operating operating mode. mode. AND OR 2. EITHER of the Deviation Adopted the revised EAL structure and V15 2. The event has caused following:
wording provided in proposed EAL FAQ VISIBLE DAMAGE to a
* Event damage has 2016-02. -SAFETY SYSTEM caused indications
+l component or structure of degraded C 0 needed for the current performance to a C"I operating mode. second train of the <t VI SAFETY SYSTEM needed for the current operating mode, or
* The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. Intent and meaning of the EALs are not altered. 77 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
SS1 SS1 Verbatim Global Comment #14 None Initiating Condition:
Loss of all Loss of all offsite and all onsite Verbatim None offsite and all onsite AC power to AC power to emergency buses for emergency buses for 15 minutes 15 minutes or longer. or longer. Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None .-I Power Operation, Startup, Hot 2, 3, 4 V) V) Standby, Hot Shutdown (1) Loss of ALL offsite and ALL (1) Loss of ALL offsite and ALL Difference Global Comment #9, 12 & 13 None onsite AC power to (site-onsite AC power to 1(2)-A05 specific emergency buses) and 1(2)-A06 for 15 minutes for 15 minutes or longer. or longer. Intent and meaning of the EALs are not altered. 78 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
555 556 Verbatim Renumbered to align with other similar !Cs None Initiating Condition:
Inability to Inability to shutdown the reactor Difference Global Comment #4 None shutdown the reactor causing a causing a challenge to core challenge to (core cooling [PWR] cooling or RCS heat removal. / RPV water level [BWR]) or RCS heat removal. Operating Mode Applicability:
Operating Mode Applicability:
1 Difference Global Comment #10 None Power Operation (1) a. An automatic or manual (1) a. An automatic or manual Difference Global Comment #4, 9 & 12 None (trip [PWR] / scram [BWR]) trip did not shutdown the did not shutdown the reactor. reactor. AND AND Verbatim None in b. All manual actions to b. All manual actions to V) V) shutdown the reactor have shutdown the reactor been unsuccessful.
have been unsuccessful.
AND AND Difference Global Comment #9 V19 & V20 C. EITHER ofthe following
: c. EITHER of the following conditions exist: conditions exist: * (Site-specific indication
* Conditions requiring of an inability to entry into Core Cooling adequately remove heat -Red path (CSP-C.1) from the core) are met. * (Site-specific indication
* Conditions requiring of an inability to entry into Heat Sink-adequately remove heat Red path (CSP-H.l) are from the RCS) met. Intent and meaning of the EALs are not altered. 79 PBNP DEVIATIONS AND DIFFERENCES MATRIX Sect.ion NEI 99-01 Rev .. 6 PBNP Change Justification Validation
# Recognition Category:
SS8 SS2 Verbatim Global Comment #14; renumbered to align None with other emergency power source !Cs Initiating Condition:
Loss of all Loss of all Vital DC power for 15 Verbatim None Vital DC power for 15 minutes or minutes or longer. longer. Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3, 4 co Standby, Hot Shutdown V, V, (1) Indicated voltage is less (1) Indicated voltage is less Difference Global Comment #9 & 12 V33 than (site-specific bus than 115 VDC on ALL voltage value) on ALL (site-Vital DC busses D-01, D-specific Vital DC busses) for 02, D-03, and D-04 for 15 minutes or longer. 15 minutes or longer. Intent and meaning of the EALs are not altered. 80 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
SG1 SGl Verbatim Global Comment #14 None Initiating Condition:
Prolonged Prolonged loss of all offsite and Verbatim None loss of all offsite and all onsite AC all onsite AC power to emergency power to emergency buses. buses. Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None Power Operation, Startup, Hot 2,3,4 Standby, Hot Shutdown (1) a. Loss of ALL offsite and ALL (1) a. Loss of ALL offsite and ALL Difference Global Comment #9 & 13 None onsite AC power to (site-onsite AC power to 1(2)-specific emergency buses) . ADS and 1{2)-A06.
.... Difference C, AND AND Global Comment #9 & 13 V34 VI b. EITHER of the following:
: b. EITHER of the following:
* Restoration of at least
* Restoration of at least one AC emergency bus one AC emergency bus in less than (site-specific in less than 4 hours is hours) is not likely. not likely. * (Site-specific indication
* Conditions requiring Difference Global Comment #9 V19 of an inability to entry into Core Cooling adequately remove heat -Red Path {CSP-Cl) from the core) are met. Intent and meaning of the EALs are not altered. 81 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section ,NEI 99:-01 Rev. 6 PBNP Change Justification Validation
# Recognition Category:
SG8 SG2 Verbatim Global Comment #14; renumbered to align None with other emergency power source ICs Initiating Condition:
Loss of all AC Loss of all AC and Vital DC power Verbatim None and Vital DC power sources for sources for 15 minutes or longer. 15 minutes or longer. Operating Mode Applicability:
Operating Mode Applicability:
1, Difference Global Comment #10 None Power Operation, Startup, Hot 2, 3, 4 Standby, Hot Shutdown (1) a. Loss of ALL offsite and ALL (1) a. Loss of ALL offsite and ALL Difference Global Comment #9 & 12 None co onsite AC power to (site-onsite AC power to 1(2)-V'l specific emergency buses) A05 and 1(2)-A06 for 15 for 15 minutes or longer. minutes or longer. AND AND Difference Global Comment #9 & 13 V33 b. Indicated voltage is less b. Indicated voltage is less than (site-specific bus than 115 VDC on ALL Vital voltage value) on ALL (site-DC busses D-01, D-02, D-specific Vital DC busses) for 03, and D-04 for 15 15 minutes or longer. minutes or longer. Intent and meaning of the EALs are not altered. 82 PBNP DEVIATIONS AND DIFFERENCES MATRIX APPENDIX A-ACRONYMS AND ABBREVIATIONS 83 PBNP DEVIATIONS AND DIFFERENCES MATRIX *section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# AC. ...... Alternating Current AC. ...... Alternating Current Verbatim N/A AOP ...... Abnormal Operating AOP ...... Abnormal Operating Verbatim N/A Procedure Procedure APRM ... Average Power Range Difference Not used N/A Meter ATWS ... Anticipated Transient ATWS ... Anticipated Transient Verbatim N/A Without Scram Without Scram B&W .... Babcock and Wilcox Difference Not used N/A BIIT ...... Boron Injection Initiating Difference Not used N/A Temperature Vl BWR .... Boiling Water Reactor Difference Not used N/A z 0 CDE ...... Committed Dose CDE ...... Committed Dose Verbatim N/A > Equivalent Equivalent w CFR ...... Code of Federal CFR ...... Code of Federal Verbatim N/A c::: cc Regulations Regulations cc <( CTMT/CNMT
... Containment CNMT ... Containment Difference CTMT not used N/A C z CSF ...... Critical Safety Function CSF ...... Critical Safety Function Verbatim N/A <( Vl CSFST ... Critical Safety Function CSFST ... Critical Safety Function Verbatim N/A > Status Tree Status Tree z DBA ......
Design Basis Accident Difference Not used 0 N/A er:: u DC. ....... Direct Current DC. ....... Direct Current Verbatim N/A <( I EAL. ...... Emergency Action Level EAL. ...... Emergency Action Level Verbatim N/A <( ECCS .... Emergency Core Cooling ECCS .... Emergency Core Cooling Verbatim N/A X c System System z ECL. ...... Emergency Classification ECL. ...... Emergency Classification Verbatim w N/A C. C. Level Level <( EOF ...... Emergency Operations EOF ...... Emergency Operations Verbatim N/A Facility Facility EOP ...... Emergency Operating EOP ...... Emergency Operating Verbatim N/A Procedure Procedure EPA ...... Environmental Protection EPA ...... Environmental Protection Verbatim N/A Agency Agency EPG ..... Emergency Procedure Difference Not used N/A Guideline EPIP ..... Emergency Planning Difference Not used N/A Implementing Procedure 84 PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. Ei PBNP Change Justification Validation#
EPR ...... Evolutionary Power Difference Not used N/A Reactor EPRI. .... Electric Power Research Difference Not used N/A Institute ERG ..... Emergency Response Difference Not used N/A Guideline FEMA ... Federal Emergency FEMA. .. Federal Emergency Verbatim N/A Management Agency Management Agency FSAR .... Final Safety Analysis Difference Not used N/A -Report ...; s:: GE ........ General Emergency GE ........ General Emergency Verbatim N/A 0 HCTL....Heat Capacity Difference Not used N/A II') 2 Temperature Limit 0 HPCI... .. High Pressure Coolant Difference Not used N/A > Injection w HSI. ....... Human System Interface Difference Not used N/A CZ:: cc cc IC. .........
lnitiating Condition IC. .........
lnitiating Condition Verbatim N/A <( C ID .........
lnside Diameter ID .........
lnside Diameter Verbatim N/A z IPEEE ... lndividual Plant Difference Not used N/A <( II') Examination of External Events > (Generic Letter 88-20) z 0 ISFSl. ... lndependent Spent Fuel ISFSl. ... lndependent Spent Fuel Verbatim N/A CZ:: u Storage Installation Storage Installation
<( I Keff. .... Effective Neutron Keff ..... Effective Neutron Verbatim N/A <( X Multiplication Factor Multiplication Factor 15 LCO ..... Limited Condition of LCO ..... Limited Condition of Verbatim N/A 2 w Operation Operation C. C. LOCA. .. Loss of Coolant Accident LOCA ... Loss of Coolant Accident Verbatim N/A <( MCR .... Main Control Room Verbatim N/A MSIV ... Main Steam Isolation Difference Not used N/A Valve MSL... .. Main Stem Line Difference Not used N/A mR, mRem, mrem, mREM .... milli-mR, mRem, mrem, mREM .... milli-Verbatim N/A Roentgen Equivalent Man Roentgen Equivalent Man MW ..... Megawatt MW ..... Megawatt Verbatim N/A NEI. ...... Nuclear Energy Institute NEI. ...... Nuclear Energy Institute Verbatim N/A NPP ...... Nuclear Power Plant Difference Not used N/A 85 PBNP DEVIATIONS AND DIFFERENCES MATRIX ... Section NEI 99-01 Rev. 6 *. PBNP Change Justification Validation
# NRC. .... Nuclear Regulatory NRC. .... Nuclear Regulatory Verbatim N/A Agency Agency NSSS .... Nuclear Steam Supply Difference Not used N/A System NORAD ... North American NORAD ... North American N/A Aerospace Defense Command Aerospace Defense Command (NO)UE ... (Notification of) Unusual (NO)UE ... (Notification of) Unusual Verbatim N/A Event Event -...: NUMARC. ... Nuclear Management NUMARC. ... Nuclear Management Verbatim N/A C: 0 and Resources Council and Resources Council V) OBE ..... Operating Basis OBE ..... Operating Basis Verbatim N/A 2 0 Earthquake Earthquake OCA ..... Owner Controlled Area OCA ..... Owner Controlled Area Verbatim N/A > ODCM/ODAM
.... Offsite Dose ODCM ... Offsite Dose Calculation Difference PBNP uses ODCM N/A w c::: cc Calculation (Assessment)
Manual Manual cc c:( ORO ..... Offsite Response ORO ..... Offsite Response Verbatim N/A C Organization Organization z c:( PA. ........ Protected Area PA .........
Protected Area Verbatim N/A V) 2: PACS .... Priority Information and Difference Not used N/A > 2 Control System 0 c::: PAG ...... Protective Action PAG ...... Protective Action Verbatim N/A u c:( Guideline Guideline I c:( PICS ..... Process Information and Difference Not used N/A X Control System l5 2 PRA/PSA ... Probabilistic Risk PRA/PSA. .. Probabilistic Risk Verbatim N/A w C. Assessment/Probabilistic Safety Assessment/Probabilistic Safety C. c:( Assessment Assessment PWR .... Pressurized Water Reactor PWR .... Pressurized Water Reactor Verbatim N/A PS .........
Protection System Difference Not used N/A PSIG .... Pounds per Square Inch PSIG .... Pounds per Square Inch Verbatim N/A R ..........
Roentgen R ..........
Roentgen Verbatim N/A RCC. ... Reactor Control Console Difference Not used N/A RCIC. .. Reactor Core Isolation Difference Not used N/A Cooling 86 PBNP DEVIATIONS AND DIFFERENCES MATRIX Sectiqn NEI 99-01* Rev: 6 PBNP Change Justification Validation
# RCS ..... Reactor Coolant System RCS ..... Reactor Coolant System Verbatim N/A Rem, rem, REM ... Roentgen Rem, rem, REM ... Roentgen Verbatim N/A Equivalent Man Equivalent Man RETS .... Radiological Effluent Difference Not used N/A Technical Specifications RPS ...... Reactor Protection System RPS ...... Reactor Protection System Verbatim N/A RPV ...... Reactor Pressure Vessel RPV ...... Reactor Pressure Vessel Verbatim N/A -..; RVLIS ... Reactor Vessel Level RVLIS ... Reactor Vessel Level Verbatim N/A C: 0 Instrumentation System Instrumentation System II) RWCU ... Reactor Water Cleanup Difference Not used N/A z 0 SAR ....... Safety Analysis Report Difference Not used N/A SAS ........ Safety Automation Difference Not used N/A > w System ci:: CD SBO ....... Station Blackout Difference Not used N/A CD <( SCBA ..... Self-Contained Breathing SCBA ..... Self-Contained Breathing Verbatim N/A C z Apparatus Apparatus
<( SG ..........
Steam Generator SG ..........
Steam Generator Verbatim N/A II) 2: SI. ..........
Safety Injection SI... ........ Safety Injection Verbatim N/A > z SICS ...... Safety Information Difference Not used N/A 0 ci:: Control System u <( SPDS ..... Safety Parameter Display SPDS ..... Safety Parameter Display Verbatim N/A I <( System System X Difference c SRO ....... Senior Reactor Operator Not used N/A z TEDE ..... Total Effective Dose TEDE ..... Total Effective Dose Verbatim N/A w c.. c.. Equivalent Equivalent
<( TOAF ..... Top of Active Fuel TOAF ..... Top of Active Fuel Verbatim N/A TSC. ....... Technical Support TSC.. ...... Technical Support Verbatim N/A System System -UFSAR .... Final Safety Analysis Difference Used in Section 3.1 N/A Report WOG ..... Westinghouse Owners WOG ..... Westinghouse Owners Verbatim N/A Group Group 87 PBNP DEVIATIONS AND DIFFERENCES MATRIX APPENDIX B -DEFINITIONS 88 PBNP DEVIATIONS AND DIFFERENCES MATRIX IC NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Alert: Events are in progress or have occurred Alert: Events are in progress or have occurred Verbatim None which involve an actual or potential which involve an actual or potential substantial degradation of the level of safety substantial degradation of the level of safety of the plant or a security event that involves of the plant or a security event that involves probable life threatening risk to site probable life threatening risk to site personnel or damage to site equipment personnel or damage to site equipment because of HOSTILE ACTION. Any releases are because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of expected to be limited to small fractions of the EPA PAG exposure levels. the EPA PAG exposure levels. V, General Emergency:
Events are in progress or General Emergency:
Events are in progress or Verbatim None z have occurred which involve actual or have occurred which involve actual or 0 E IMMINENT substantial core degradation or IMMINENT substantial core degradation or z u:::: melting with potential for loss of melting with potential for loss of w C containment integrity or HOSTILE ACTION containment integrity or HOSTILE ACTION I i:a that results in an actual loss of physical that results in an actual loss of physical X control of the facility.
Releases can be control of the facility.
Releases can be 15 z reasonably expected to exceed EPA PAG reasonably expected to exceed EPA PAG w c.. exposure levels offsite for more than the exposure levels offsite for more than the c.. <( immediate site area. immediate site area. Notification of Unusual Event: Events are in Unusual Event: Events are in progress or have Difference See Global Comment #3 None progress or have occurred which indicate a occurred which indicate a potential potential degradation of the level of safety of degradation of the level of safety of the plant the plant or indicate a security threat to or indicate a security threat to facility facility protection has been initiated.
No protection has been initiated.
No releases of releases of radioactive material requiring radioactive material requiring offsite offsite response or monitoring are expected response or monitoring are expected unless unless further degradation of safety systems further degradation of safety systems occurs. occurs. 89 PBNP DEVIATIONS AND DIFFERENCES MATRIX IC NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Site Area Emergency:
Events are in progress Site Area Emergency:
Events are in progress Verbatim None or have occurred which involve actual or or have occurred which involve actual or likely major failures of plant functions needed likely major failures of plant functions needed for protection of the public or HOSTILE for protection of the public or HOSTILE ACTION that results in intentional damage or ACTION that results in intentional damage or malicious acts; 1) toward site personnel or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure equipment that could lead to the likely failure of or; 2) that prevent effective access to, of or; 2) that prevent effective access to, equipment needed for the protection of the equipment needed for the protection of the public. Any releases are not expected to public. Any releases are not expected to result in exposure levels which exceed EPA result in exposure levels which exceed EPA V'I PAG exposure levels beyond the site PAG exposure levels beyond the site z boundary.
boundary.
0 E Emergency Action Level (EAL): A pre-Emergency Action Level (EAL): A pre-Verbatim None z determined, site-specific, observable determined, site-specific, observable i:i: w threshold for an Initiating Condition that, threshold for an Initiating Condition that, C I when met or exceeded, places the plant in a when met or exceeded, places the plant in a co X given emergency classification level. given emergency classification level. 15 z Emergency Classification Level (ECL): One of a Emergency Classification Level (ECL): One of a Difference See Global Comment #3 None w c.. set of names or titles established by the US set of names or titles established by the US c.. <( Nuclear Regulatory Commission (NRC) for Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions grouping off-normal events or conditions according to (1) potential or actual effects or according to (1) potential or actual effects or consequences, and (2) resulting onsite and consequences, and (2) resulting onsite and offsite response actions. The emergency offsite response actions. The emergency classification levels, in ascending order of classification levels, in ascending order of severity, are: severity, are:
* Notification of Unusual Event (NOUE)
* UnusualEvent(UE)
* Alert
* Alert
* Site Area Emergency (SAE)
* Site Area Emergency (SAE)
* General Emergency (GE)
* General Emergency (GE) 90 PBNP DEVIATIONS AND DIFFERENCES MATRIX IC NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Fission Product Barrier Threshold:
A pre-Fission Product Barrier Threshold:
A pre-Verbatim None determined, site-specific, observable determined, site-specific, observable threshold indicating the loss or potential loss threshold indicating the loss or potential loss of a fission product barrier. of a fission product barrier. Initiating Condition (IC): An event or Initiating Condition (IC): An event or Verbatim None condition that aligns with the definition of condition that aligns with the definition of one of the four emergency classification one of the four emergency classification levels by virtue of the potential or actual levels by virtue of the potential or actual effects or consequences.
effects or consequences.
CONFINEMENT BOUNDARY: (Insert a site-CONFINEMENT BOUNDARY:
The barrier(s)
Difference Removed developer notes None specific definition for this term.) Developer between spent fuel and the environment and added site-specific Note -The barrier(s) between spent fuel and once the spent fuel is processed for dry language.
the environment once the spent fuel is storage. This corresponds to the pressure processed for dry storage. boundary for the Dry Shielded Canister (DSC) in NU HOMS 32PT casks and the Multi-assembly Sealed Basket (MSB) for the VSC-24 Storage System. CONTAINMENT CLOSURE: {Insert a site-CONTAINMENT CLOSURE: Is the action taken Difference Removed developer notes None specific definition for this term.) Developer to secure containment and its assorted and added existing Note -The procedurally defined conditions structures, systems and components as a definition from present or actions taken to secure containment functional barrier to fission product release EALs. V) (primary or secondary for BWR) and its under existing plant conditions.
2 associated structures, systems, and CONTAINMENT CLOSURE is initiated per the 0 E components as a functional barrier to fission SEPs or Shift Manager direction if plant 2 u: product release under shutdown conditions.
conditions change that could raise the risk of w a fission product release as a result of a loss C I co of decay heat removal. CONTAINMENT X CLOSURE requires that, upon a loss of decay c 2 heat removal, any open penetration which is w c... listed on CL lE, Containment Closure c... <t Checklist, must be closed or capable of being closed prior to RCS bulk boiling. This checklist is maintained any time that the RCS is <200&deg;F and CONTAINMENT operability is not maintained.
91 PBNP DEVIATIONS AND DIFFERENCES MATRIX I IC NEI 99-01 Rev. 6 PBNP Change Justification I Validation
# I EXPLOSION:
A rapid, violent and catastrophic EXPLOSION:
A rapid, violent, and catastrophic Verbatim None failure of a piece of equipment due to failure of a piece of equipment due to combustion, chemical reaction or combustion, chemical reaction, or overpressurization.
A release of steam (from overpressurization.
A release of steam (from high energy lines or components) or an high energy lines or components) or an electrical component failure (caused by short electrical component failure (caused by short circuits, grounding, arcing, etc.) should not circuits, grounding, arcing, etc.) should not automatically be considered an explosion.
automatically be considered an explosion.
Such events may require a post-event Such events may require a post-event II) inspection to determine if the attributes of an inspection to determine if the attributes of an z explosion are present. explosion are present. 0 E z FAULTED: The term applied to a steam FAULTED: The term applied to a steam Difference Removed developer note. U::: None w generator that has a steam leak on the generator that has a steam leak on the C I secondary side of sufficient size to cause an secondary side of sufficient size to cause an cc X uncontrolled drop in steam generator uncontrolled drop in steam generator iS z pressure or the steam generator to become pressure or the steam generator to become w Cl. completely depressurized.
Developer Note -completely depressurized.
Cl. <( This term is applicable to PWRs only. FIRE: Combustion characterized by heat and FIRE: Combustion characterized by heat and Verbatim None light. Sources of smoke such as slipping drive light. Sources of smoke such as slipping drive belts or overheated electrical equipment do belts or overheated electrical equipment do not constitute FIRES. Observation offlame is not constitute FIRES. Observation of flame is preferred but is NOT required if large preferred but is NOT required if large quantities of smoke and heat are observed.
quantities of smoke and heat are observed.
HOSTAGE: A person(s) held as leverage HOSTAGE: A person(s) held as leverage Verbatim None against the station to ensure that demands against the station to ensure that demands will be met by the station. will be met by the station. 92 PBNP DEVIATIONS AND DIFFERENCES MATRIX IC . NEI 99-01 Rev. 6 PBNP Change Justification Validation
# HOSTILE ACTION: An act toward a NPP or its HOSTILE ACTION: An act toward a nuclear Difference Spelled out 'NPP' in 2 None personnel that includes the use of violent power plant or its personnel that includes the places force to destroy equipment, take HOSTAGES, use of violent force to destroy equipment, and/or intimidate the licensee to achieve an take HOSTAGES, and/or intimidate the end. This includes attack by air, land, or water licensee to achieve an end. This includes using guns, explosives, PROJECTILEs, vehicles, attack by air, land, or water using guns, or other devices used to deliver destructive explosives, PROJECTILEs, vehicles, or other force. Other acts that satisfy the overall devices used to deliver destructive force. intent may be included.
HOSTILE ACTION Other acts that satisfy the overall intent may should not be construed to include acts of be included.
HOSTILE ACTION should not be civil disobedience or felonious acts that are construed to include acts of civil not part of a concerted attack on the NPP. disobedience or felonious acts that are not Non-terrorism-based EALs should be used to part of a concerted attack on the nuclear address such activities (i.e., this may include power plant. Non-terrorism-based EALs violent acts between individuals in the owner should be used to address such activities (i.e., controlled area). this may include violent acts between individuals in the owner controlled area). HOSTILE FORCE: One or more individuals who HOSTILE FORCE: One or more individuals who Verbatim None are engaged in a determined assault, overtly are engaged in a determined assault, overtly or by stealth and deception, equipped with or by stealth and deception, equipped with suitable weapons capable of killing, maiming, suitable weapons capable of killing, maiming, or causing destruction.
or causing destruction.
IMMINENT:
The trajectory of events or IMMINENT:
The trajectory of events or Verbatim None conditions is such that an EAL will be met conditions is such that an EAL will be met within a relatively short period of time within a relatively short period of time regardless of mitigation or corrective actions. regardless of mitigation or corrective actions. INDEPENDENT SPENT FUEL STORAGE INDEPENDENT SPENT FUEL STORAGE Verbatim None INSTALLATION (ISFSI): A complex that is INSTALLATION (ISFSI): A complex that is designed and constructed for the interim designed and constructed for the interim storage of spent nuclear fuel and other storage of spent nuclear fuel and other radioactive materials associated with spent radioactive materials associated with spent fuel storage. fuel storage. 93 PBNP DEVIATIONS AND DIFFERENCES MATRIX IC NEI 99-01 Rev. 6 PBNP Change Justification Validation
# NORMAL LEVELS: As applied to radiological Difference Term not used in this EAL None IC/EALs, the highest reading in the past scheme twenty-four hours excluding the current peak value. OWNER CONTROLLED AREA: (Insert a site-OWNER CONTROLLED AREA: The site Difference Definition from developer None specific definition for this term.) Developer property owned by or otherwise under the notes used. Developer Note -This term is typically taken to mean control of the licensee.
Notes deleted. the site property owned by, or otherwise under the control of, the licensee.
In some cases, it may be appropriate for a licensee to define a smaller area with a perimeter closer to the plant Protected Area perimeter (e.g., a ti) site with a large OCA where some portions of z 0 the boundary may be a significant distance E z from the Protected Area). In these cases, u:: developers should consider using the w C boundary defined by the Restricted or I cc Secured Owner Controlled Area X c (ROCA/SOCA).
The area and boundary z w selected for scheme use must be consistent
: a. a. with the description of the same area and <C boundary contained in the Security Plan. PROJECTILE:
An object directed toward a NPP PROJECTILE:
An object directed toward a Difference Spelled out 'NPP' None that could cause concern for its continued nuclear power plant that could cause concern operability, reliability, or personnel safety. for its continued operability, reliability, or personnel safety. PROTECTED AREA: (Insert a site-specific PROTECTED AREA: The area under Difference Definition from developer None definition for this term.) Developer Note-continuous access monitoring and control, notes used. Developer This term is typically taken to mean the area and armed protection as described in the site Notes deleted. under continuous access monitoring and Security Plan. control, and armed protection as described in the site Security Plan. 94 PBNP DEVIATIONS AND DIFFERENCES MATRIX IC .. NEI 99-01 Rev. 6 PBNP Change Justification Validation #
* REFUELING PATHWAY: (Insert a site-specific REFUELING PATHWAY: The reactor refueling Difference PBNP-specific definition None definition for this term.) Developer Note-cavity, spent fuel pool, and fuel transfer supplied.
Developer This description should include all the canal. Notes deleted. cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. RUPTURE(D):
The condition of a steam RUPTURE(D):
The condition of a steam Difference Removed developer None generator in which primary-to-secondary generator in which primary-to-secondary notes. leakage is of sufficient magnitude to require a leakage is of sufficient magnitude to require a safety injection.
Developer Note -This term safety injection.
is applicable to PWRs only. II) z SAFETY SYSTEM: A system required for safe SAFETY SYSTEM: A system required for safe Difference Removed developer notes None 0 j:: plant operation, cooling down the plant plant operation, cooling down the plant and clarified last sentence.
2 u: and/or placing it in the cold shutdown and/or placing it in the cold shutdown w condition, including the ECCS. These are condition, including the ECCS. These systems C I typically systems classified as safety-related.
are classified as safety-related.
c:a X Developer Note -This term may be modified c z to include the attributes of "safety-related" in w C. accordance with 10 CFR 50.2 or other site-C. <( specific terminology, if desired. SECURITY CONDITION:
Any Security Event as SECURITY CONDITION:
Any Security Event as Verbatim None listed in the approved security contingency listed in the approved security contingency plan that constitutes a threat/compromise to plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or site security, threat/risk to site personnel, or a potential degradation to the level of safety a potential degradation to the level of safety of the plant. A SECURITY CONDITION does of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION. not involve a HOSTILE ACTION. SITE BOUNDARY:
That line beyond which the Difference Defined term from ODCM None land is neither owned, nor leased, nor needed for several EALs otherwise controlled by the licensee.
95 PBNP DEVIATIONS AND DIFFERENCES MATRIX NEl99-Ql Rev. 6. l?BNP. Change J . , Justificatio.n
! 'Validation
#.J UNISOLABLE:
An open or breached system UNISOLABLE:
An open or breached system Verbatim None line that cannot be isolated, remotely or line that cannot be isolated, remotely or locally. locally. UNPLANNED:
A parameter change or an UNPLANNED:
A parameter change or an Verbatim N/A event that is not 1) the result of an intended event that is not 1) the result of an intended evolution or 2) an expected plant response to evolution or 2) an expected plant response to a transient.
The cause of the parameter a transient.
The cause of the parameter V) change or event may be known or unknown. change or event may be known or unknown. 2 0 E VISIBLE DAMAGE: Damage to a component or VISIBLE DAMAGE: Damage to a component or Deviation Updated to reflect VlS 2 structure that is readily observable without structure that is readily observable without wording and guidance of u::: w measurements, testing, or analysis.
The visual measurements, testing, or analysis.
The visual proposed EAL FAQ 2016-C I impact of the damage is sufficient to cause a:i impact of the damage is sufficient to cause 02. The updated wording X concern regarding the operability or concern regarding the operability or clarifies damage 15 2 reliability of the affected component or reliability of the affected component or assessment meriting an w D. structure.
structure.
Damage resulting from an ALERT declaration as used D. <( equipment failure and limited to the failed in !Cs using this definition component (i.e., the failure did not cause (CA6 and SA9). damage to a structure or any other equipment) is not VISIBLE DAMAGE. 96 PBNP DEVIATIONS AND DIFFERENCES MATRIX APPENDIX C -Permanently Defueled ICs/EALs PBNP DEVIATIONS AND DIFFERENCES MATRIX Section NEI 99-01 Rev. 6 PBNP Change Justification Validation
# Appendix C -Permanently Not used at PBNP Difference Not applicable to PBNP None > Defueled ICs/EALs +l C: QJ !I <[ E w ... -QJ Ill Cl.~ I "C u .!!! QJ .?5 :i "C .... C: QJ QJ C 0.. 0.. <[ 98 ATTACHMENT 4 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 286, ADOPTION OF EMERGENCY ACTION LEVEL SCHEME PURSUANT TO NEI 99-01 REVISION 6, , "DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR NO.N-PASSIVE REACTORS" UPDATED SUPPORTING TECHNICAL INFORMATION 189 pages follow MODE TITLE 1 Power Operation 2 Startup 3 Hot Standby 4 Hot Shutdown(b) 5 Cold Shutdown(b) 6 Refueling(c) (a) Excluding decay heat. Table 1.1-1 (page 1 of 1) MODES REACTIVITY
% RATED CONDITION THERMAL (keff) POWER(a) 0.99 >5 0.99 s5 < 0.99 NA < 0.99 NA < 0.99 NA NA NA Definitions
 
===1.1 AVERAGE===
REACTOR COOLANT TEMPERATURE (OF) NA NA 350 350 > Tavg > 200 s 200 NA (b) All reactor vessel head closure bolts fully tensioned. (c) One or more reactor vessel head closure bolts less than fully tensioned.
Point Beach 1.1-7 Unit 1 -Amendment No. 201 Unit 2 -Amendment No. 206 INTE RIM STAFF G UI DANCE EMERGENCY PLANNING FOR NUCLEAR POWER PLANTS licensee to promptly dec l are the emergency condition as soon as possib l e following the identification of the a p propriate ECL. As used h ere, " promp tly" means the n ext ava il ab l e opportu n ity unimpeded by activities not related to the emerge n cy declarat i on , unless such activities are necessary for protecting hea l th and safety. (See Paragraph 8 of this section.) 6. Consiste n t with the N RC's position t hat emerge n c y declarati o ns are made p romptly , t h e final ru l e states that t h e 15-m i n u te criterion not b e construed as a grace period in which a licensee may attempt to restore plant conditions to avoid dec l aring an EAL that has already been exceeded. This statement does not preclude licensees from acting to correct or mitigate an off-normal condition , bu t once an EAL has been recognized as being exceeded , the emergency de cl aration s h a ll be made promptly without w aiting for the 15-m inute perio d to e l apse. Thi s is particu l arly the case w h en the EAL t hreshold is exceeded based on occurrence of a condition , rather than the duration of a condition. 7. For EAL thresholds that specify a duration of the off-normal condition , the NRC expects that the emergency declaration process run concurrently with the specified threshold duration. Once the off-normal condition has existed for the duration specified in the EAL , no further effort on this declaration is necessary-the EAL has been exceeded. Consider as an example , the EAL " fire which is not extinguished within 15 minutes of detection." On receipt of a fire alarm, the plant fire brigade is dispatched to the scene to begin fire suppression efforts.
* If the fire brigade reports that the fire can be extinguished before the specified duration , the emergency declaration is placed on hold while firefighting activities continue.
If the fire brigade is successful in extinguishing the fire within the specified duration from detection , no emergency declaration is warranted based on that EAL.
* If the fire is still burning after the specified duration has elapsed , the EAL is exceeded , no further assessment is necessary , and the emergency declaration would be made promptly.
As used here, " promptly" means at the first available opportunity (e.g., if the Shift Manager is receiving an update from the fire brigade at the 15-minute mark , it is expected that the declaration will occur as the next action after the call ends).
* If, for example , the fire brigade notifies the shift supervision 5 minutes after detection that the brigade itself cannot extinguish the fire such that the EAL will be met imminently and cannot be avoided , the NRC would not consider it a violation of the licensee's emergency plan to declare the event before the EAL is met (e.g., the 15-minute duration has elapsed). While a prompt declaration would be beneficial to public health and safety and is encouraged , it is not required by regulation.
* In all of the above, the fire duration is measured from the time the alarm , indication , or report was first received by the plant operators. Validation or confirmation establishes that the fire started as early as the time of the alarm , indication, or report. N SI R/DPR-ISG-01 Rev. 0 (November 2 0 11)
I I I C a l c ul at i o n 2013-0 01 8 Revis i o n 2 The mon i t or r e ading s th at cor re s pond to t h e AUl initiating condition (2X ODCM Alarm S etpoints) under NEI 99-0 1 R e v i sion 4 are liste d below: Mon i tor Rea din g RE-315, Auxiliary Building Exha ust Low Range Gas 5.4E-03 &#xb5;Ci/cc RE-317, Auxiliary Buildin g Exhaust Low Ran ge Gas 5.4E-03 &#xb5;Ci/cc 1 (2)RE-307 Containment Purge Exha ust Mid-Range Gas, with 1.4E-02 &#xb5;Ci/cc only containment purge in operation (25,000 cfm) 2RE-307, Containment Purge Exhaust Mid-Range Gas with both 9.4E-03 &#xb5;Ci/cc purge and GS building ventilation in operation (38,000 cfm). 2RE-305, Containment Purge Exhaust Low-Range Gas with both 9.4E-03 &#xb5;Ci/cc purge and GS building ventilation in operation (38,000 cfm). 2RE-307, Containment Purge Exha ust Mid-Range Gas with only 2.8E-02 &#xb5;Ci/cc GS building ventilation in operation (13 ,000 cfm). 2RE-307, Containment Purge Exhaust Mid-Range Gas with only 1.0E+Ol &#xb5;Ci/cc forced vent of containment (35 cfm). 2RE-309, Containment Purge Exha ust High Range Gas with only 1.0E+Ol &#xb5;Ci/cc forced vent of containment (35 cfm). RE-325, Drumming Area Exhaust Low-Range Gas 8.4E-03 &#xb5;Ci/cc RE-327, Drumming Area Exhaust Mid-Range Gas 8.4E-03 &#xb5;Ci/cc Page 23 o f 23 ODCM POINT BEACH NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL Revision 19 TOTAL REWRITE OFFSITE DOSE CALCULATION MANUAL TABLE 9-1 LIQUID EFFLUENT PATHWAYS LIQUID EFFLUENT PATHWAY PATHWAY MONITOR 3 Recirculation Water None Service ater eturn (n o rmal cool 1(2)RE-229 down per o@) Steam Generator 1 (2)RE-219*
Blowdown & 1 (2)RE-222 Waste Water RE-230 Effluent' Spent Fuel Pool RE-220* Waste Distillate
& RE-218* & Condensate Storage RE-223* Tank Discharge Containment Fan 1(2)RE-216*
Cooler Return DISCHARGE FLOWRATE (GPM) ............................
..... 1 pump, .. either .. unit ....................
243 , 000 ............
............
.. 2pumps , eitherunit . }94 , 000 .....................
.................
lpump, eachunit.... ..... 48:!_, 000 1 pump , one unit & 619 , 000 ..............................
2 pumps,other
.. unit .... 2 pumps , each unit 744,000 2pumps@7500gpm ..... )5 , 000. 3 pumps@ .. 6300gpm . _ 18 , 900 ___ 4~umps@ .. 510~gpm. )0 , 400 ....... 5pumps@4300_gpm )1 , 500 6 pumps @, 3 700 gpm 22 , 200 Max Flow Rate Max Flow Rate (both filter skids running in parallel)
Max Flow Rate Max Flow Rate Max Flow Rate (per Containment) 200 700 700 100 4000 CALCULATED DEFAULT SETP0INT 1 (uCi/cc) N I A N I A N I A N I A N I A 1.1;:iE-03 1.26E-01 1.22E-03 3.61E-02 2.53E-01 6.32E-03 NOTE 1: Setpoints except for RE-230 are based on lOx the MEC values listed in 10CFR20, Appendix B, Table 2 , Column 2. PBNP TS Section 5.5.4.b allows concentrations of radioactive material released to unrestricted areas to be lOx the MEC values. NOTE 2: RE-230 setpoint explanation can be found in section 9.1, Default Monitor Setpoints.
NOTE 3: Monitors marked with an asterisk(*)
have a calculated default alarm setpoint above the monitors fail high or saturation level. See section 9.1, High Alarm or Trip Setpoint Guidelines for further explanation.
2 times 1.14E-03 = 2.28E-03 Value rounded to 2.3E-03 for use in EAL RULi Page 50 of 278 REFERENCE US E Document Information:
CALCULATION COVER SHEET (Page 1 of 1). Calculation (Doc) No: 2013-0018 . I Controlled Documents Revision:
2 Title: RADIOLOGICAL EFFLUENT INITIATING CONDITION VALVES FOR EMERGENCY ACTION LEVEL RS 1 AND RG 1 Type: CALC Sub-Type:
CALC Discipline:
Radiological I Unit: 0 Facility:
PBNP Safety Class: D SR Quality Related D Non-Nuclear Safety D Important to Safety D Not Important to Safety Special Codes: D Safeguards D Proprietary Vendor Doc No: NIA I Vendor Name or Code:N/ A Executive Summary (optional):
Expand scope to include EAL ICs for Unusual Events and Alerts, and for both Revision 4 and Revision 6 ofNEI 99-01. Delete calculation of setpoints based on a 4 hour assumed release duration (make consistent with NEI 99-01 assumption of 1 hour). Incorporate inputs from ODCM Revision 19. . Review and Approval:
Associated EC Number: 288117 EC Revision-!"'0-*1.fc<.
;,j-,, 6/r:; AR/ Other Document Number: 02179019 Description of Calculation Revision:
Update EC Document Revision:
2 Prepared by: .,,,/ (Li, /&#xa3;,/7 T. C. Kendall Date: 2/16/2017 (signawre) (print.name)
Reviewed by: (1 I) A:. lL_. Carl Onesti Date: 2/16/2017
--. ,-(signature) (print name) Type of Review:~ Design Verification D Review D Owner Acceptance Review Method Used (ForDV~nly):
D Design Review D Alternate Calculation Approved by: C'\;;;,,~p.J ff'*"**------
---h--e. --2..)'2...b I Date: Z-/c;,.p/;
'----4-7 , /7 ( signafure) 1 \}-A'f'-re
"""'-(print name) EN-AA-100-1004-F01, Revision 0 I C alcu l a ti o n 2 0 13-00 1 8 R ev i s ion 2 R es ul ts and Co n cl u s i ons: T h e results were ca l culated us ing a source term t h at is co n si s tent with the O D CM and the gu i dance of NEI 9 9-01. The resu l ts are app li ca b le for a reac t or thermal po w er operat in g l e v el of 18 00 MW th and a r e a ppli c a ble genera lly to other t h e r ma l power o p erating level s due to the use of activity fract i onal data and not absolute activity values. T h e mon i to r read ing s that corres po nd to the AGl i nitiating condition value of 1 rem TED E or 5 rem T h y ro id CDE are li s t ed below. Mo nito r Reading 1(2)RE-309, Containment Purge Exhaust High Range Gas with only 600 &#xb5;Ci/cc conta i nment purge in operation (25,000 cfm) 2RE-309, Containment Purge Exhaust High Range Gas with both 400 &#xb5;Ci/c c purge and GS Building in operation (38 , 000 cfm) 2RE-309, Containment Purge Exhaust High Range Gas with only 1200 &#xb5;Ci/cc GS Building in operation (13,000 cfm) RE-317, Auxiliary Building Exhaust Mid-Range Gas 100 &#xb5;Ci/cc RE-319, Auxiliary Building Exhaust High Range Gas 100 &#xb5;Ci/cc RE-327, Drumming Area Exhaust Mid-Range Gas Off s c ale high* 1 (2) R E-231 , Steam Li ne 1A(2 A); 1 (2)R E-232 , S te a m Line l(A)2(A) Atmosp h e r ic Dump Va lv e (AD V) r e l ease 19 &#xb5;Ci/c c Main Stea m S a fety Va lv e (M SSV) re l ease 7.4 &#xb5;Ci/c c *T he u pper ra n ge lim i t on R-32 7 is 10 0 &#xb5;Ci/cc, so t his instrume n t wou l d be ov er-ranged EAL I C of 160 &#xb5;C i/cc. However , this is close en ou gh to the u pper range of the instrumen by t h e nom in a l t t h at it is judged approp r ia t e to , in the a b se n ce of real-time dose projection capab i lity and no oth make a classificat io n declarat i o n based on a va li d off-scale hi gh reading of this in st rume -EAL Developer gwdance from 99 01. er indications , t o nt. It is recognized that the condition described by this IC may result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. In those cases, EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available.
For example, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding, if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading, then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold.
Since the upper range of RE-327 is 100 &#xb5;Ci/cc and this instrument would be severely ranged by the calculated value of 160 &#xb5;Ci/cc (160% of range), this monitor will not be included in this EAL. P age 19 of 23 I Calculation 2013-00 1 8 Revision 2 T h e mo nitor readi ng s th at c orr es pon d to t h e A Sl initiati ng c on d i tion va l ue of 100 mre m TE DE or 5 re m T hyr o i d CDE are l ist e d be l ow for a one hour re l ease duration.
Monitor Readin g 1 (2)RE-307 Containment Purge Exhaust Mid-Range G as, with only 60 &#xb5;Ci/cc containment purge in operation (25,000 cfm) 1(2)RE-309 Containment Purge Exhaust High-Range Ga s, with only 60 &#xb5;Ci/cc containment purge i n operation (25,000 cfm) 2RE-307, Containment Purge Exhaust M i d Range Gas with both 40 &#xb5;C i/cc purge and GS building ventilation in operation (38,000 cfm). 2RE-309 , Containment Purge Exhaust High Range Ga s with both 40 &#xb5;Ci/cc purge and GS building ventilation in operation (38 , 000 cfm). 2RE-309 , Containment Pu rg e Exhaust High Rang e Ga s w i th only 120 &#xb5;Ci/cc GS build i ng ventilat i on i n opera t ion (13 , 000 cfm). RE-317, Auxi li ary B u ilding Exhaust Mid-Range Gas 10 &#xb5;Ci/cc RE-319, Auxi l iary Building Exhaust High Range Gas 1 0 &#xb5;Ci/cc RE-327, Drummi n g Area Exha u st Mid-Range Gas 1 6 &#xb5;Ci/c c 1(2) RE-231 , Steam Line 1A(2A); 1(2)RE-232 , Steam Line l(A)2(A) Atmospheric Dump Valve (ADV) release 1.9 &#xb5;Ci/cc Main Steam Safety Va l ve (MSSV) re l ease 0.74 &#xb5;Ci/cc P age 2 0 of 23 I I I I Calc ul at i on 2013-0 01 8 Revision 2 T h e mo nitor rea din g s that cor r es p ond to t h e AA l i nitiating con d ition (1/10 0 of AGl) u n der N EI 99-01 e ndm ent Re vi s i on 6 are li s ted b e l ow. The s e will r e qu ire s u bmitta l an d ap p rov a l of a License Am Requ est to im plem e n t. Monitor Readin g 1(2)RE-307 Containment Purge Exhaust Mid-Range Gas, with only 6.0 &#xb5;Ci/cc containment purge in operation (25,000 cfm) 1(2)RE-309 Containment Purge Exhaust High-Range Gas, with only 6.0 &#xb5;Ci/c c containment purge in operation (25 ,000 cfm) 2RE-307, Containment Purge Exhaust Mid Range Gas with both 4.0 &#xb5;Ci/cc purge and GS building ventilation in operation (38 , 000 cfm). 2RE-309, Containment Purge Exhaust High Range Gas with both 4.0 &#xb5;Ci/cc purge and GS building ventilation in operation (38,000 cfm). 2RE-309, Containment Purge Exhaust High Range Gas with only 1 2 &#xb5;Ci/cc GS building ventilation in operation (13,000 cfm). RE-317, Auxiliarv Buildin g Exha ust Mid-Range Gas 1.0 &#xb5;Ci/cc RE-319, Auxiliarv Building Exha ust High Range Gas 1.0 &#xb5;Ci/cc RE-327, Drumming Area Exhaust Mid-Range Gas 1.6 &#xb5;Ci/cc 1(2) RE-231 , Steam Line 1A(2A); 1(2)RE-232 , Steam Line l(A)2(A) Atmospheric Dump Va l ve (ADV) release 0.19 &#xb5;Ci/cc Main Steam Safety Valve (MSSV) release 0.074 &#xb5;Ci/cc P age 21 of 23 I I I Calculation 2013-0018 Revision2 The monitor readings that correspond to the AA1 initiating condition (200X ODCM Alar m Setpoints) under NEI 99-01 Revision 4 are listed below: Monitor Reading RE-315, Auxiliary Building Exhaust Low Range Gas 5.4E-01 &#xb5;Ci/cc RE-317, Auxiliary Building Exhaust Low Range Gas 5.4E-01 &#xb5;Ci/cc 1(2)RE-307 Containment Purge Exhaust Mid-Range Gas, with l.4E+OO &#xb5;Ci/cc only containment purge in operation (25,000 cfm) 1(2)RE-309 Containment Purge Exhaust High-Range Gas, with l.4E+OO &#xb5;Ci/cc only containment purge in operation (25,000 cfm) 2RE-307, Containment Purge Exhaust Mid-Range Gas with both 9.4E-01 &#xb5;Ci/cc purge and GS building ventilation in operation (38,000 cfm). 2RE-307, Containment Purge Exhaust Mid-Range Gas with only 2.8E+OO &#xb5;Ci/cc GS building ventilation in operation (13,000 cfm). 2RE-309, Containment Purge Exhaust High Range Gas with only 2.8E+OO &#xb5;Ci/cc GS building ventilation in operation (13,000 cfm). 2RE-309, Containment Purge Exhaust High Range Gas with only l.OE+03 &#xb5;Ci/cc forced vent ofcontainment (35 cfm). RE-327, Drumming Area Exhaust Mid-Range Gas 8.4E-01 &#xb5;Ci/cc Page 22 of 23 I I I I Ca l cu l at i on 2013-0018 R ev i s i on 2 The mon itor r eading s th at c orre s pon d to th e AUl in i tiat i ng cond ition (2X ODCM Alar m S etpoi nts) under NEI 99-0 1 R e vi s i on 4 a r e li s ted below: Mo n itor Rea din g RE-315, Auxiliarv Building Exhaust Low Range Gas 5.4E-03 &#xb5;Ci/cc RE-317, Auxiliary B u ilding Exha u st Low Range Gas 5.4E-03 &#xb5;Ci/cc 1(2)RE-307 Containment Purge Exhaust Mid-Range Gas , with 1.4E-02 &#xb5;Ci/cc on l y containment purge in operation (25 , 000 cfm) 2RE-307 , Containment Purge Exhaust Mid-Range Gas with both 9.4E-03 &#xb5;Ci/cc purge and GS bui l ding venti l ation in operation (38,000 cfm). 2RE-305, Containment Purge Exhaust Low-Range Gas with both 9.4E-03 &#xb5;Ci/cc purge and GS bui l ding ventilation in operation (38 , 000 cfm). 2RE-307 , Containment Purge Exhaust Mid-Range Gas with only 2.8E-02 &#xb5;Ci/cc GS building venti l ation in operation (13 , 000 cfm). 2RE-307, Containment Purge Exhaust Mid-Range Gas with only l.OE+Ol &#xb5;Ci/cc forced vent of containment (35 cfm). 2RE-309, Containment Purge Exhaust High Range Gas with only 1.0E+Ol &#xb5;Ci/cc forced vent of containment (35 cfm). RE-325 , Drumming Area Exhaust Low-Range Gas 8.4E-03 &#xb5;Ci/cc RE-327, Drumming Area Exhaust Mid-Range Gas 8.4E-03 &#xb5;Ci/cc Page 2 3 of 2 3 ENERCON Excellence
-Every project Every day 1. Purpose and Scope CALC. NO. Fuel Handling Accident Monitor Response for REV. NEE-363-CALC-002 0 EAL Thresholds
------------1 PAGE NO. 6 of 23 The purpose of this calculation is to determine the expected dose rates on radiation monitors RE-126 , RE-127, RE-128, and RE-135 during a fuel handling accident (FHA) at Point Beach Nu cl ear Plant (PBNP). Monitors RE-126 , RE-127 , and RE-128 are located in the reactor containment building (RCB). Monitor RE-135 is inside the Auxiliary (AUX) Bui l ding near the Spent Fuel Poo l (SFP). The accident occurs either in the SFP or RCB 408 hours after shutdown. The results are used as thresho l d values for Emergency Action Level (EAL) RA2.2 in the PBNP EAL Technical Basis document , which implements NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors". The containment building , the auxiliary building , and components within the bu i ldings are modeled simplistically because only order of magnitude results are needed. As such , the dose rate results should be considered as reasonably representative of the magnitude of the actual dose rate only. This calcu l ation is not Nuclear Safety Related as the results of the calculation does not affect the design basis or Safety Related systems structures or components. This calculat i on represents an as built analysis of plant conditions , therefore no acceptance criteria is required.
: 2. Summary of Results and Conclusion The results of this calculation are listed below. Table 2-1 Detector Resoonse Location Monitor Dose Rate (R/h) RE-126 5.73 RCB RE-127 5.57 RE-128 5.57 SFP RE-135 3.84 Reading levels at or above the values listed in Table 2-1 will be indicative of a fuel handling accident.
The dose rates reported do not include the background (ambient) radiation readings associated with the monitor calibration.
For monitors 126, 127, and 128, a normal background level of 1.5 R/hr was added to the calculated value, and the resulting values were then rounded to whole numbers as follows for use in EAL RA2.2. RE-126 = 7 R/hr RE-127 = 7 R/hr RE-128 = 7 R/hr SFP monitor RE-135 does not have an existing background reading of significance, so the calculated value was rounded as follows for use in EAL RA2.2: RE-135 = 4 R/hr 
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Local Operations for Normal Operation/Shutdown/Cooldown Step-by-step analysis was performed on the following PBNP procedures to determine those " rooms or areas t ha t contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation , coo/down and shutdown":
* OP 3A: Power operation to hot standby
* OP 38: Reactor Shutdown
* OP 3C: Hot Standby to Cold Shutdown
* OP ?A: Placing RHR System in operation
* OP 50 Part 4: Degassing the RCS using the PZR and Letdown Gas Stripper Analysis did not include rooms or areas i n which actions of a contingent o r emergency nature would be performed. (e.g., an action to address an off-normal or emergency condition such as emergency repairs , corrective measures or emergency operations). Analysis also did not i nclude rooms or areas for which entry is requ i red solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
Backup to CR or Vital to Procedure Procedure:
Mode: Step: Action: automatic plant Location of vital action: action? ops OP3A 5.2.3 1 Start MFP seal water pump s y N OP3A 5.4.4 1 SW overboard alignment N N OP3A 5.5.3 1 shut blast damper N N OP3A 5.9.3 1 Open MSR purge valves N N OP3A 5.11.1 1 Bypass LP Feed heater coolers N N OP 3 A 5.11.5 1 shut mov-1 and 2 y N OP 3A 5.13.5 1 start turbine bearing l ift o il y N pumps OP3A 5.20.4 1 lube oil cooldow n n N OP 3 8 n one n/a n/a n/a n/a Procedure Backup to CR or Vital to Procedure:
Mode: Action: automatic plant Location of vita l action: Step: action? ops OP 3C 5.1.1 3/4/5 Put the flex pumps in 8' fan N N room OP 3C 5.1.7 3/4/5 Sample blender output N y VCT Area OP 3C 5.1.8 3/4/5 degas the RCS N y See OP SD P4 below OP 3C 5.6 3/4/5 N2 to the VCT N y VCT Area OP3C 5.12 4/5 Isolate accumulators N y C-59 area OP3C 5.13 4 align flex accumulato r N N OP3C 5.16 4 Isolate SI pump y N ba ck up action s to CR only OP 3C 5.23.6 s Align conta i nment purge N N OP 3 C 5.23.7 4 align fle x N2 N N Restra i nt for entry into Mode Sonly , plant is stable and can remain in this condition until area accessible OP 3C attachment A 3/4/5 borate to refueling y N backup to CR only concentration Stop at step 5.24 because plant is now in mode 5. -
Procedure Backup to CR or Vital to Procedure:
Mode: Action: automatic plant Location of vital action: Step: action? ops OP 7A 5.l.3a 4 adjust CC cooling to RHR N y Pipeway 2/3, 8' elev. OP 7A 5.1.4 4 caution tag 851 871s N y C-59 area OP 7A 5.1.5 4 align RHR suction N y C-59 area, -Sft OP7A 5.1.6 4 shut RH-716 N N OP 7A 5.2.16 4, 5 CCW temp control N y CCW HX Room OP SD P4 5.2 3 In i tiate primary degas N y primary sample room OP SD P4 5.3.3 3 tag shut CV-261c N y VCT area OP SD P4 5.3.5 3 isolate H2 to vet N y VCT area Resulting Tables used in EALs RA3 and HAS are shown below: Table R-2 SAFE OPS, S/D, C/D AREAS Area/Building MODE Ul VCT Area 3/4/5 U2 VCT Area 3 /4 / 5 Ul Primary Sample area 3 U2 Primary Sample area 3 CCW HX Room 4/5 C-59 area 3/4/5 Pipeway 2, 8 ft. Elev. 3/4 Pipeway 3, 8 ft. Elev. 3/4 1/2B-32 MCC Area 4 Table H-2 SAFE OPS, S/D, C/D AREAS Area/Building MODE Ul VCT Area 3/4/5 U2 VCT Area 3 /4 I 5 Ul Primary Sample area 3 U2 Primary Sample area 3 CCW HX Room 4/5 C-59 area 3/4/5 Pipeway 2 , 8 ft. Elev. 3/4 Pipeway 3, 8 ft. Elev. 3/4 1/2B32 MCC Area 4 
 
===6.5 LEAKAGE===
DETECTION SYSTEMS Leakag e Detection Systems FSAR Section 6.5 The leak detection systems reveal the presence of significant leakage from the reactor coolant , residual heat remova l , and component cooling systems. 6.5.l DESIGN BASIS Monitoring Reactor Coolant Leakage Criterion:
Means shall be provided to detect significant uncontrolled l eakage from the reactor coolant pressure boundary. (GDC 16) Positive indications in the control room of leakage of coolant from the reactor coolant system to the containment are provided by equipment which permits continuous monitoring of containment air activity and humidity , and of runoff from the air recirculation units and containment floor drains to containment Sump A. his equipment provides indication of normal background which is indicative of a basic level of leakage from p r imary systems and components.
Any increase in the o b served parameters is an indication of change within the containment , and the equipment provided is capable of monitoring this change. The basic design criterion is the detection of deviations from normal containment environmental conditions including air particulate activity , gaseous activity, humidity , condensate and floor drain runoff and , in ad d ition , in the case of gross leakage , the liquid inventory in the process systems and containment sump. See Section 15.4.3 for additional informat i on regarding leak detection requirements for leak-before-break analyses. Criterion:
Means shall be provided for monitoring the containment atmosphere and the facility effluent discharge paths for radioactivity released from normal operations , from anticipated transients, and from accident conditions. An environmental monitoring program shall be maintained to confirm that radioactivity releases to the environs of the plant have not been excessive. (GDC 17) The following are monitored for radioactivity concentrations during normal operation, anticipated transients, and acc i dent conditions:
the containment atmosphere , the exhausts from the 54 in. auxiliary and serv i ce bui l ding vent, the 46 in. d rumming area vent, the 36 in. containment area vents, the 4 in. combined air ejector exhaust vent, the service water discharge from the containment fan coolers , the component cooling loop liquid, the liquid phase of the secondary side of the steam ge n erator, waste disposal system liquid disc h arge, spent fue l pool heat exchanger service water return , waste distillate discharge , gas stripper building ventilation exhaust, service water discharge, wastewater effluent, steam line atmospheric release, and the con d enser air ejector. GDC 17 is a l so addressed in Section I 1.4, Radiation Protection. A continuing environmental monitoring program, discussed in Section 2.0, is maintained.
Principles of Design The principles for d esign of the leakage detection systems can be summarized as follows: I. Increased leakage could occur as the result of failure of pump seals, valve packing g l ands , flange gaskets, or instrument connections.
The maximum leakage rate calculated for these types of failures is 50 gpm which would be the anticipated flow rate of water through the pump seal if the entire sea l were wiped out and the area between the shaft and housing were completely open. UFSAR2015 Page 6.5-1 of 10 
======---------
------* u AC Sources-Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources-Operating LCO 3.8.1 The following AC eleclrlcal power sources shall be OPERABLE:
: a. One circuit between lhe off site transmission network and the associated unit's 4.16 kV Class 1E safegua rd s buses, A05andA06, utilizing the associated unit's 345/13.8 kV (X03) transformer or the opposite unit's 345/13.8 l<V (X03) transformer with the gas turbine In operation, and the associated unit's 13.8/4 .16 kV (X04} transformer;
: b. One circuit between the offsite transmission network and the opposite unit's 4.16 kV Class 1E safeguards buses, A05 and A06; and c. One standby emergency power source capable of supplying each 4.16 kV/480 V Class 1 E safeguards bus. -APPLICABILITY:
MODES 1, 2, 3, and 4. ACTIONS -----------------------------------------------------NOTE-------------------------------------------
LCO 3.0.4.b is not applicable to standby emergency power sources. CONDITION A. Associated unit A.1 345/13.8 kV (X03) transformer inoperable. .QR Gas turbine not In operation when utilizing opposite unit's 345/13.8 kV (X03) transformer.
AMO. Point Beach REQUIRED ACTION C0\1PLETION11ME Verify one circuit 24 hours between the off site transmission network and the associated unit's 4.16 kV Class 1 E safeguards buses, A05 and A06, utilizing the opposite unit's 345/13.8 kV (X03) transformer. (continued) 3.8.1-1 Unit 1 -Amendment No. 21 5 Unit 2 -Amendment No. 22 E C B z 0 f-:::J QJ 0:: f-Vl 0 0:: w t_'"*' Q_ ~* _J <l: u 0:: u w _J w Q_ z co Q_ 5 A 09lt, 2 UPDATE PRI: 2 OCR: YES S I M?\.IFIEO 
 
===3.8 ELECTRICAL===
 
POWER SYS T EMS 3.8.4 DC Sources-Operating DC Sources-Operating 3.8.4 LCO 3.8.4 The D-01, D-02, D-03, and D-04 DC e l ectrical powe r subsystems shall be OPERABLE. APPLICABILITY
: MODES 1, 2 , 3 , and 4. ACTIONS CONDITION REQUIRED ACTION A. One DC electrical power ---------------
--NOTE----------------subsystem inoperable. Enter applicable Conditions and Required Act i ons of LCO 3.8.9 , " Distribution Systems-Operating ," when any DC bus i s de-energized. ------------------------------------------A.1 Restore DC electrical power subsystem to OPERABLE status. B. Required Action and B.1 Be in MODE 3. Associated Completion Time not met. AND B.2 Be in MODE 5. SURVEILLANCE REQUIREMENTS SR 3.8.4.1 SURVEILLANCE Verify correct battery t erm i nal voltage is within limits on float charge. COMPLETION TIME 2 hours 6 hours 36 hours FREQUENCY In accordance w i th the Surveillance Frequency Control Program Point Beach 3.8.4-1 (continued)
Unit 1 -Amendment No. 253 Unit 2 -Amendment No. 257 POINT BEACH NUCLEAR PLANT SETPOINT DOCUMENT EMERGENCY OPERA TING PROCEDURE (EOP) SETPOINTS PNBP ERG STPT STPT F.18 Y.01 F.19 F.20 L.07 F.21 F.22 L.11 F.23 L.12 F.24 Y.02 F.25 Y.03 F.26 Y.04 G.l M.01 G.2 M.02 PARAMETER Rx Vessel Level sh utdown Rx Vessel Level (shutdown)
Rx Vessel Level Rx Vessel Level Rx Vessel Level Rx Vessel Level Rx Vessel Level Rx Vessel Level SG Level (narrow range) SG Level (narrow range) DESCRIPTION in imum shutdown r eactor vessel l evel for o eration of RAR without air bind ing tne suction Shutdown reactor vessel level range for indication that level is on scale Wide range level corresponding to a void fraction of25% with both RCPs running, including normal channel accuracy SPARE Wide Range reactor vessel level corresponding to a void fraction of 0% with both RCPs running Wide Range reactor vessel level corresponding to a void fraction of 0% with one RCP running Shutdown reactor vessel level corresponding to mid-loop elevation in the RCS hot legs , including allowances for normal channel accuracies.
Shutdown reactor vessel level corresponding to just onscale in the RCS hot legs , including allowances for normal channel accuracies. Shutdown reactor vessel level corresponding to the top of the RCS hot legs , including allowances for normal channel accuracies. Steam generator narrow range no-load level Steam generator narrow range level indicating U-tubes covered, including normal channel accuracy and process errors Page 9 of28 STPT 25.l Revision 14 SETPOI T Unit l Unit 2 16% 16% 10% to 90% 10% to 90% 65 ft 65 ft 85 ft 85 ft 45 ft 45 ft 20% 20% 4% 4% 33% 33% 64% 64% 32% 32% CALC REFERENCE 2010-0027 EOPSTPT Doc F.19 2010-0018
 
2010-0018 2010-0018 2010-0018 2010-0027 2010-0027 2010-0027 2010-0019 2010-0019 
 
Lake Level Hydrology FSAR Section 2.5 The nominal water level in Lake Michigan at the time of the original license submittal was -2 feet relative to the Plant Datum. A maximum water level was recorded in 1886 at+ I. 7 feet and minimum recorded to date occurred in 1964 at-4.8 feet. The site is , on average, about 20 or more feet above plant elevation zero and there is no record that it has been flooded by the lake. The maximum analyzed value for high lake level is+ 1.7 feet. Operators will take actions to commence the orderly shutdown of any operating reactor per Abnormal Operating Procedure direction prior to reaching the analyzed limit. Flood Level T h e license basis level for rotection of critical equi ment from lake flooding is +9.0 feet (Reference 25). This is an acceptable and bounding value as each lake flooding source when evaluated individually, or in the combined effects review, provides resultant flood levels conservatively below this threshold thereby satisfying the General Design Criteria 2 requirement to include "an appropriate margin for withstanding forces greater than recorded to reflect uncertainties about the historical data and their suitability as a basis for design." The limiting design basis lake flood event is a combination of the maximum lake level, the maximum wave run-up and a conservative value for the wind setup effect. Details are provided in the following section entitled "Combined Effects." The calcu l ated level reaches +7.25 feet on the riprap shoreline and +8.42 feet on the vertical surfaces of the intake structure (Reference 28). All critical plant components are therefore protected by the strategies outlined in Appendix A. 7 "Plant Flooding ,"(Reference
: 30) . Tides on Lake Michigan created by the attraction of the moon and sun are insignificant.
The total range of oscillation does not exceed 2 inches. Using the method delineated in "The Prediction of Surges in the Southern Basin of Lake Michigan, Part 1, The Dynamical Basis for Prediction" by G. W. Paltzman (Reference 31 ), the storm surge that could occur at the site will be 4.14 feet due to the passage of a squall line with a pressure jump of 8 millibars and a simultaneous speed of movement of 65 knots with a shoaling factor of 3 .5. Adding this surge of 4.14 feet to the maximum recorded water level in Lake Michigan of+ 1. 7 feet results in a maximum elevation of 5 .84 feet, which is bounded by the license basis flood level and is considerably lower than the turbine building grade floor elevation of +8.0 feet or the pumphouse operating floor elevation of +7 .0 feet. The va l ue of 4.14 feet was developed using Pla t zman's contours of amplitude for pressure.
There are no contours for the lake in the area of the site so a conservative approach was taken using the reflected surge values for Waukegan at 90&deg; with a speed of movement of 65 knots, giving a pressure rise of 0.05 feet. Using 8 millibars or 0.236", the maximum surge due to pressure with a 3.5 shoaling factor will, therefore, be UFSAR2015 Page 2.5-4 of 15 
 
====3.7.7 Service====
Water (SW) System Service Water (SW) System TRM 3.7.7 TLCO 3. 7. 7 The SW System shall be operated in accordance with the evaluated conditions and alignments of an analyses demonstrating the ability to provide required cooling water flow to required equipment. APPLICABILITY:
MODES 1, 2 , 3 and 4. ACTIONS --------------------------
----------------------
-NOTE----------------------------------------------------
Performance of an evaluation, within the required Completion Time , that verifies the SW System is capable of providing required cooling water flow to required equipment is an acceptable alternative to the stated Required Actions. CONDITION A SW Intake temperature
> 85&deg;F. B Pump Bay level < -11.5 feet. C Pump Bay level< -15 feet POINT BEACH TRM A.1 B.1 C. 1 REQUIRED ACTION COMPLETION T I ME Declare G-01, G-02, Immediately Containment Fan Coolers, PAS Battery Room Vent Coolers, Component Cooling Water (CCW) heat exchangers (HX), inoperable. Declare Unit 1 " A", Immediately Unit 2 "A", and Unit 2 "B" accident fan cooler units inoperable. Declare G01 and G02 Immed i ately inoperable and all containment fan coolers inoperable. (continued) 3.7.7-1 November 2015 Determination of SW Pump Minimum Submergence Calculation P-89-037, Revision 2 RESULTS AND CONCLUSIONS Page 8 The SW pumps will perform adequately with a minimum pumpbay water level at the -19 (ft) elevation and an individual pump flowrate up to 7800 (gpm). A water level below the -19 (fi) elevation could result in vortexing due to inadequate inlet submergence. Cavitation should not occur since adequate NPSH will be available down to the -26 (ft) elevation.
Since the pump inlet is located at the -26 (ft) elevation, the pump would become airbound prior to cavitation. Pumpbay water level is far superior to pump discharge pressure for indication of the onset of inadequate pump performance. Adequate pump performance is a function of both submergence and flowrate inputs (flowrate is used directly in Equation (7) and indirectly in Equation (6) to determine NPSHR). Discharge pressure alone cannot serve as an indicator of cavitation because the fluctuating pumpbay level is not accounted/or.
Since minimum submergence has been determined for a conservatively high flowrate, pumpbay level should be used in lieu of discharge pressure to directly indicate the onset in inadequate pump performance. The following documents and actions have been initiated by this calculation.
* Engineering Work Request (EWR)-Since the discharge pressure sctpoint is not based on cavitation prevention, an EWR will be initiated to evaluate the proper source for the sctpoint.
* Potential Open Item For DBD Program (PBF-I 611) -Service Water Design Basis Document
* Procedure Feedback Request (PBF-0026p) 70, Rev. 31, Sec. 3.13 will be revised to reflect the independence of cavitation prevention from discharge pressure setpoint.
* Training Department Notification
-Training Reference Handbook (TRHB) 11.8, Rev.7, Sec. 5.2.3 should be revised to reflect the independence of cavitation prevention from discharge pressure.
Emergency Preparedness Program Frequently Asked Question (EPFAQ) EPFAQ Number: Originator:
Organization:
Relevant Guidance:
2016-002 . David Young NEI NEI 99-01, Methodology for Development of Emergency Action Levels, Revisions 4 and 5; and NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 6. NUMARC/NESP-007, Methodology for Development of Emergency Action Levels. Applicable Section(s):
Initiating Condition (IC) HA2 in NEI 99-01, Revisions 4 and 5, and NUMARC/NESP-007, "FIRE or EXPLOSION Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown" Status: NOTE: ICs CA6 and SA9 in NEI 99-01, Revision 6: "Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode" Definition of VISIBLE DAMAGE in NEI 99-01, Revisions 4, 5 and 6, and NUMARC/NESP-007 Available for Public Comment Based on industry comments provided by Jetter dated February 16, 2017 (ADAMS Accession No. ML 17079A228), and subsequent staff discussions, a proposed revision to these /Cs was proposed in the public meeting held on April 4, 2017, and was attached to the public meeting notice (ADAMS Accession No. ML 17089A458).
Based on comments provided by the industry during the April 4, 2017 public meeting, the staff revised the proposed revisions to these /Cs. QUESTION OR COMMENT: A review of industry Operating Experience has identified a need to clarify an aspect of the definition of VISIBLE DAMAGE as it relates to the I Cs cited above; adding this clarity is necessary to minimize the potential for an over-classification of an equipment failure. There may be cases where VISIBLE DAMAGE is the result of an equipment failure and limited to the failed component (i.e., the failure did not cause damage to any other component or a structure).
The current definition of VISIBLE DAMAGE does not adequately differentiate between damage resulting from, and affecting only, the failed piece of equipment vs. an equipment failure causing damage to another component or a structure (e.g., by a failure-induced fire or explosion).
Can the definition of VISIBLE DAMAGE be clarified to help avoid an inappropriate emergency declaration in cases where an equipme.nt failure does not result in damage to another component or a structure (i.e., VISIBLE DAMAGE affects only the failed component)?
A related question is also posed -Consistent with the approach used. in other I Cs, should a note be added to preclude an emergency declaration if the safety system affected by a hazard was not functional before the event occurred (e.g., tagged out for maintenance)?
PROPOSED SOLUTION:
Yes; the sentence below may*be added to the definition of VISIBLE DAMAGE [as defined in NEI 99-01, Revisions 4, 5, and 6]. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. From a plant safety and change-in-risk perspective, the consequences from the failure of a piece of equipment, accompanied by a hazard (e.g., a fire or explosion) that does not damage 1 Emergency Preparedness Program Frequently Asked Question (EPFAQ) any other equipment or a structure , are essentially the same as the equipment fai l i ng w i th no atte nd a n t haza r d. N either eve nt w o uld appear to meet t h e defin i tion of an Alert b ecause the outcome does not involve an a ctu al or pote n tial substant i a l degradation of the leve l of safety o f the pl ant (e.g., there has been n o significant reduction in t h e margin to a l oss or pote n tial loss of a fission produc t barrier).
N uc l ear power p l ants are designed with redundant safety system trains that are re q uired to be se p arated (i.e., installed in separate plant areas or have sepa ra t i on within an individua l a r ea). Absent any collateral damage to another component or a structure , a hazard associated with an equi p ment failure does not affect the ability to protect publ i c health and safety , and there is no addit io nal response benefit to be gained by declaring an emergency. The norma l plant organization has sufficient resources and adequate guidance to respond to an equi p ment fail ure -guidance incl udes ope r a t ing proced u res and Tec h nical Specifications
; the fi re prot e ct io n [progra m], industria l safety and corrective act i o n programs; a n d work ma n agement and m aintenance re q uirements. Concerning the second questi o n , an emergency declaration would not be appropriate in resp on se to a hazard affecting a piece of equipment or s y stem that was non-functiona l prior to the event (e.g., ta g ged out for m aintenance). For this reason and consistent with t h e approach use d in other ICs, the following n ote may be added to IC H A2 (NEI 99-0 1 R4 and R5), or ICs CA6 and SA9 (NE I 99-01 R6). Note: If the affected safety system (or component) was already non-functional befo r e the event occurred, then no emergency clas s ification is warranted. Co n s i st e nt with t h e g uidance i n Re g ulatory I ssue Summar y (RIS) 2003-18 , Supp l eme nt 2 , Use of N u cl ear Energy I nstitute (N E/) 99-01, " M et h odology for D evelopment of Emergenc y Action Levels ," Revision 4 , dated Ja n uary 2003 , it is reasonab l e to conclude t h at the changes proposed above would be considered as a " deviation." NRC RES P ONSE: The p roposed gui d ance is inte nd e d to ensure that an Alert s h ould be dec l ared on ly when actual or pote n tial performance issues w ith SAFETY SYSTEMS have occurred as a resu l t of a haza rd ous event. The occurrence of a hazardous event w ill result in a Notification of Unusual Eve n t (NOUE) c l assification at a minimum. In order to warrant escalation to the Alert class i fication , the h azardous e v e n t should ca u se indicat ion s of degra d ed performance to one tra in of a SAFET Y SYSTEM w ith e it h er in d ica t ions of deg r aded performance on t h e second SA F E T Y SYSTE M t r a i n or V I S IBLE DA MA GE to the sec on d SAFET Y SYSTEM tra in , such tha t the o perability or re l iability of t h e second trai n is a concern. In addition , escalation t o the Alert class i fication shou l d not occur i f the damage from the hazardous event is limited to a SAFETY SYSTEM that was inoperable , o r out of seN i ce , prior to the event occurring. As such , the proposed guidance will reduce the potential of declaring an Alert when events are in progress that d o not involve a n actual or po tential substantial degradation of the level of safety of the pla n t , i.e., does no t cause sig nifi c an t conce rn w i th shutting down or coo l ing down t h e pl ant. IC HA2 (NEI 99-0 1 R4 and R 5; NU MARC/N ESP-007), o r ICs CA6 and SA9 (NEI 99-01 R6), do not directly escalate to a Site Area Emergency or a General Emergency due to a hazardous event. The Fission Product Barrier and/or Abnormal Radiation Levels/Radiological Effluent recognition categories would provide an escalation path to a Site Area Emergency or a General Emergency. The proposed addition of the following notes , applicable to ICs HA2 (NE I 99-01 R4 an d R5; NUMARC/NESP-00 7), or ICs CA6 and SA9 (NEI 99-01 R6), provide further clarification as to how these Alert emergency classifications are cons i dered. The revisions to these EALs , including the addition of the notes , are consistent with the current NRC-endorsed Alert 2 Emergency Preparedness Program Frequently Asked Question (EPFAQ) classification l anguage. 1. Adding th e following no t e to the applicable EALs , per this E P F AQ , is acceptable as it meets the intent of the EALs , is consistent with other EA L s (e.g., EAL HAS from NEI 99-01, Revision 6; this revision was endorsed by the NRC in a letter dated March 28 , 2013 , available a t ADAMS Accession No. ML 1 2346A463), and ensures that declared emergencies a r e based upon unplanned events with the potential to pose a rad i ological risk t o the public. If t he affected safe t y syste m tr ain (or component) w as already inoperable o r out of service before the event occurred , then this emergency classification is not w a rran t ed a s long as the damage i s limited to the affected safety system train (or componen t). 2. Adding the following note to help explain the EAL is reasonable to succinctly capture the more detailed information from the Basis section related to when conditions would require the declaration of an Alert. If the event results in VISIBLE DAMAGE , w ith no ind i cations of degraded performance to any SAFETY SYSTEM train , then this emergency declaration is not warranted. Revising the EALs and the Basis sect i ons to ensu r e potential escalations from a NOUE to an Alert , due to a hazardous event , is appropriate as the concern with these EA L s is: (1) a hazardous event has occurred , (2) one SAFETY SYSTEM train i s having performance issues as a result of the hazardous event , and (3) either the second SAFETY SYSTEM train is having performance issues or the VISIBLE DAMAGE is enough to be concerned that the second SAFETY SYSTEM train may have operability or reliability issues. Revising the definition for VISIBLE DAMAGE is appropriate as th i s definit i on is only used for these EALs and the rev i sed EALs are based upon SAFETY SYSTEM trains rather than individual components or structures. All of the changes discussed above are addressed in the attached markups to NEI 99-01 , Revision 6. Licensees that use NESP-007 , NEI 99-01 Revision 4 , or NEI 99-01 Revision 5 EAL schemes can adopt this language in the relevant format the staff approved for their use. Consistent with the guidance in Regulatory Issue Summary (RIS) 2003-18 , Supplement 2 , Use of Nuclear Energy Inst i tute (NE/) 99-01 , " Methodology for Development of Emergency Action Levels ," Revision 4 , dated January 2003 , it is reasonable to conclude that the changes proposed (discussed above and as attached) would be considered as a " dev i ation." RECOMMENDED FUTURE ACTION(S):
D INFORMATION ONLY , MAINTAIN EPFAQ UPDATE GUIDANCE DURING NEXT REVISION 3 Emergency Preparedness Program Frequently Asked Question (EPFAQ) CA6 ECL: Alert Initiating Condition:
Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode. Operating Mode Applicability:
Cold Shutdown , Refueling Example Eme r gency Action Levels: Notes:
* If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred , then this emergency class i fication is not warranted.
* If the hazardous event only resulted in VISIBLE DAMAGE , with no indications of degraded performance to at least one tra i n of a SAFETY SYSTEM , then this emergency classification is not warranted. (1) Basis: a. The occurrence of ANY of the following hazardous events:
* Se i smic event (earthqua k e)
* Internal or external flood i ng event
* High winds or tornado strike
* FIRE
* EXPLOSION
* (site-specific hazards)
* Other events with similar hazard characteristics as determined by the Shift Manager AND b. 1. Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode. AND 2. EITHER of the following:
* Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode , or
* Event damage has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification , the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train , and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words , in o r der for this EAL to be classified, the hazardous event must occur , at least one SAFETY SYSTEM train must have indications of degraded performance , and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE 4 Emergency Preparedness Program Frequently Asked Question (EPFAQ) such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria 1.b.1 of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.
The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information.
This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via IC AS 1 . Developer Notes: For (site-specific hazards), developers should consider including other significant , site-specific hazards to the bulleted list contained in EAL 1.a (e.g., a seiche}. Nuclear power plant SAFETY SYSTEMS are comprised of two or more separate and redundant trains of equipment in accordance with site-specific design criteria. ECL Assignment Attributes
: 3.1.2.B 5 Emergency Preparedness Program Frequently Asked Question (EPFAQ) SA9 ECL: Alert Initiating Condition:
Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode. Operating Mode Applicability:
Power Operation , Startup, Hot Standby , Hot Shutdown Example Emergency Action Levels: Notes:
* If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then this emergency classification is not warranted.
* If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted. (1) Basis: a. The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* (site-specific hazards)
* Other events with similar hazard characteristics as determined by the Shift Manager AND b. 1. Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode. AND 2. EITHER of the following:
* Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode , or
* Event damage has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification , the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train , and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified , the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE 6 Emergency Preparedness Program Frequently Asked Question (EPFAQ) such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria 1.b.1 of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. Escalation of the emergency classification level would be via ICs FS1 or AS1. Developer Notes: For (site-specific hazards}, developers should cons i der including other significant , site-spe c ific hazards to the bulleted list contained in EAL 1.a (e.g., a seiche). Nuclear power plant SAFETY SYSTEMS are comp r ised of two o r mo r e separate and r e dundant trains of equipment in acco r dance with s i te-specific design criteria. E CL Assignment Attri b utes: 3.1.2.B 7 Emergency Preparedness Program Frequently Asked Question (EPFAQ) VISIBLE DAMAGE: Damage to a SAFETY SYSTEM train that is readily observable without measurements , testing , or analysis.
The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train. 8 
/ ENERCON ExcelJence-Every project. E v ery day. Dose Rate Evaluation of CALC NO. NEE-363-CALC-001 Reactor Vessel Water Levels,___
___________ -----, During Refueling for EAL Thresholds REV. 0 1. PURPOSE AND SCOPE The purpose of this calculation is to evaluate dose rates as a function of water height in the reactor vessel during cold shutdown or refueling operations in order to set Emergency Action Level (EAL) thresholds for core uncovery (RA2 , CS1 , CG1 ). The dose rates are calculated at the locations of the containment monitors RE-126 , RE-127 and RE-128 so that dose rate measurements by these devices can be correlated to the water level in the core , upon failure of other water level detection systems. This calculation will determine the dose rate at full core uncovery , as well as maximum water levels with a detectable dose rate response applicable to both Unit 1 and Unit 2. This calculation is not Nuclear Safety Related as the results of the calculation do not affect the design basis or Safety Related systems structures or components.
These results are best estimates based on as-built condit i ons and provide information to operators with respect to classifying an emergency , therefore no acceptance criteria is required. 2.
 
==SUMMARY==
OF RESULTS AND CONCLUSION The dose rate results for the configuration without the reactor vessel head and with the reactor vessel head are provided in Section 7.8.1 and Section 7.8.2 , respectively. The minimum dose rates with the core uncovered (i.e. water at the top of the active fuel) are shown in the table below. The dose rates reported below do not include the ambient readings associated with the monitor calibration (generally 1 to 2 R/h). Table 2-1 Dose Rate at Top of Active Fuel Model Description Dose Rate (R/h) Head Off 1.09E+02 Head On 2.94E-02 1 Detailed results of the dose rate as a function of water height are provided in Table 7-3 for cases with the head removed. Calculated value for RX head in place is below usable scale for the available radiation monitors and is therefore not used. Calculated value for RX head in place (109 R/hr) was rounded to 100 R/hr for ease of use in the EALs as an approximate indication of the underlying plant condition being assessed.
1 For the case with the head in place , the dose rate is below the detectable range of the radiation monitors of 1 R/h. Page 6 of 37 POINT BEACH NUC LEAR PLANT BACKGROUND DOCUMENT LOSS OF REACTOR OR SECONDARY COOLANT EOP STEP: 23 STEP: Check Containment Hydrogen Concentration:
PURPOSE: BG EOP-1 Revision 37 Page 38 of 52 ERG STEP: 22 To check if an excessive containment hydrogen concentration is present. BASIS: This step instructs the operator to obtain a current hydrogen concentration measurement.
Depending upon the magnitude of the hydrogen concentration, the operator will either continue with this procedure or notify the TSC to determine additional recovery actions concerning Post Ac ciden t Containment Hydrogen Reduction before continuing with the procedure. When inadequate core cooling has occurred, the containment hydrogen concentration may be as much as 10 to 12 volume percent, depending on the amount of metal-water reaction (to produce hydrogen) that has occurred in the core. The hydrogen concentration is of concern since a flammable mixture can burn, if an ignition source is available, and cause a sudden rise in containment pressure which may challenge containment integrity.
The operator is instructed to obtain a current containment hydrogen concentration measurement at this point in order to ascertain the potential flammability of the combustible gases in the containment.
Note that in order to have the potential for flammable hyd rogen concentrations, an inadequate core cooling situation must have already existed. Without an inadequate core cooling situation , sufficient hydrogen would not be expected to have been produced to cause potentially flammable mixtures.
A determination is made of the flammability of the hydrogen mixtu re with respect to the possible containment pressure rise. If the hydrogen mixture is between 0.5 volume percent and 6.0 v o lume 2ercent in dr~ air, either no hydrogen burn is possible or a limited burn may occur which does not produce a significant pressure rise. If greater than 0.5 volume percent, the operator is instructed to consult with TSC staff to determine if Post Accident Containment Hydrogen Reduction should be performed to slowly reduce containment hydrogen concentration using ventilation.
If the hydrogen concentration is less than 0.5 volume percent in dry air , a flammable situation is not imminent and the operator continues with this procedure.
HSM or HSM-H Dose Rate Evaluation Program 5.4 5.4 HSM or HSM-H Dose Rate Evaluation Program 5.4.1 The licensee shall establish a set of HSM dose rate limits which are to be applied to DSCs used at the site to ensure the limits of 1 O CFR Part 20 and 1 O CFR 72.104 are met. Limits shall establish peak dose rates at the following three locations: 1) HSM front bird screen , 2) HSM front surface , and 3) Outs ide HSM door a t the HSM door centerline. 5.4.2 Notw it hstanding the limits established in 5.4.1 , the dose rate l i mits listed below for the Standardized HSM and HSM-H shall be met when a specific DSC model loaded with fuel is stored within a module: Dose Rate Limits for the Standardized HSM and HSM-H Dose Rate Dose Rate Dose Rate Limit Limit Limit HSM Front Outside End Shield Bird Screen HSM Door Wall Exterior DSC Model HSM Model (mrem/hour) (mrem/hour) (mrem/hour) 24 P S ta ndard ized HSM 350 70 55 528 S ta ndard ize d HSM 350 70 55 61ST Standardized HSM 1300 200 15 32PT Standardized HSM 850 200 6 24PHB Standardized HSM 525 20 275 24 PTH* Standardized HSM 525 70 3 00 618TH Standardized HSM 200 100 15 24 PTH HSM-H 1300 2 5 618TH HSM-H 650 2 4 32 PTH1 HSM-H 525 2 2 698TH HSM-H 250 2 4 37PTH HSM-H 525 2 4
* Appltcable only to 24PTH-S-LC.
2 times 32 PT DSC limits (mrem/hr)
HSM Frt = 1,700 HSM door = 400 End Shield Wall Ext.= 12 The number and locations of the dose rate measurements on the surface of front bird screen of the HSM are indicated below:
* Two dose rate measurements are taken for each front bird screen for the HSM-H. These dose rate measurements are approximately within 24 inches measured from the surface of the ISFSI pad and are approximately 6 inches from the centerline of each front bird screen.
* For the standardized HSM models , three dose rates are taken on the surface of each front bird screen. The central dose location shall be at the approximate centerline of the front bird screen. The other two dose locations are spaced at approximately equal distance on either side of the central dose location.
All dose locations shall be at least 24 inches above the pad surface.
* None of these measurements shall exceed the specified dose rate limits. Standardized NUHOMS System Technical Specifications ( cont i nued) 5-13 
* *
* VSC-24 Storage Cask Final Safety Analysis Report Docket No. 72-1007 1.2.4 Maximum External Surface Dose Rate Limit/Specification:
The external surface average dose rate from all types of radiation will be less than 100 mrem per hour on the sides and 200 mrem/hr on the top. Dose rates at the air inlets and outlets will be below 350 and 100 mrem/hr, respectively. Applicability
: This dose rate limit shall apply to the entire external surface of the VCC, except the bottom surface. Objective: Action: The external dose rate is limited to this value to ensure that the cask has not been inadvertently loaded with fuel not meeting the specifications in Section 2.0 of the FSAR, to provide verification for plant personnel that radiation levels are acceptably low, and to satisfy the 10CFR72 dose rate limit of25 mrem/year at the ISFSI controlled area boundary. If the measured dose rates are above those values listed above, correct fuel loading shall be verified. If correct fuel is loaded, specific analyses must demonstrate compliance with 10 CFR Part 20 and 10 CFR Part 72 radiation protection requirements, or appropriate action must be taken to comply with the acceptable limits. A letter report , summarizing the action taken and the results of investigation conducted to determine the cause of the high dose rates, shall be submitted to the NRC within 30 days. The report must be submitted using instructions in l O CFR 72.4 with a copy sent to the administrator of the appropriate NRC regional office. Surveillance:
The external surface dose rate shall be measured after loading the MSB in the VCC and before transfer to the storage pad. The side dose rate shall be measured at a distance of 5 feet from the bottom of the VCC and at four equally spaced radial locations.
The top dose rate shall be measured at the VCC lid center and the YCC outer lid edge. The dose rate measurement shall account for the effects of background radiation on the absolute dose rate measurements . LAR 1007-006, Revi s ion 0 TS-22 2 times VSC-24 limits sides= 200 top= 400 inlets= 700 outlets = 200 June 2005 POINT BEACH NUCLEAR PLANT CRITICAL SAFETY PROCEDURE CSP-ST.0 UNIT 1 SAFETY RELATED Revision 9 CRITICAL SAFETY FUNCTION STATUS TREES Page 6 of 10 FIGURE 2 ST-2 CORE COOLING Note: Blockage of the core inlet due to debris p a ss i ng through the sump screen may occur w hen on cont a inment sump recirculation. Significant core blockag* can b* lndlcat*d by Core Exit Thermocouple& (CETs) Ind i cating greater than saturation when RVLIS shows t he core to be covered. If this occurs , RVLI S Indications of water levenn the luel are no t accurate. Procedure transitions should be based on CET trends (i.e. if CETs are trending up , then the higher level CSP is entered; i f CETs are trending down , then the lo w er leve l CSP ls entered. CORE EXI T !C s L ESS fl l,\N 1 ill0 <f N O Y E S AT LI: tlE R CP RUN NUI G RCS SUBC O LI N BA.SE O ON CORE EXJT T C~ GREATER n w,.i 17-l'f) 35*f N O YE S COA E E>O T !Cs L ESS I HAN700'f NO VE N O Y E S REACTOR VESSE L NAA fiOW RANC E LE'vl: L GREATER THAN 1 4 FT AE/, C T O A V ES S EL NARR O W RAI i E LE V E L GAEAl ER THAii 1 4 fT R E/\C T OR VESSEL WOE 11M *ELEV L GR l;A I E R rtt r..N ~n. TW O R CP, 4 0 FT, O N E R C P GOTO CSP-C.1 GOTO CSP.C.1 r JO YES GOTO CSP-C.2 GOTO CSP-C.2 N O YES GOTO CSP&deg;C.3 GOTO CSP&deg;C.2 NO YE GOTO CSP*C.3 ~----o~~~
POINT BEACH NUCLEAR PLANT CRITICAL SAFETY PROCEDURE CSP-ST.0 UNIT 1 SAFETY RELATED Revision 9 CRITICAL SAFETY FUNCTION STATUS TREES TOTAL FIGURE 3 ST-3 HEAT SINK FEEDWATER NO FLOWTOS/Gs GREATER THAN OR EQUAL TO YES 230GPM NO NARROW NO RANGE LEVEL PRESSURE IN IN ANY S/G BOTH S/Gs GREATER THAN LESS THAN [51%)32% YES 1----------11105PSIG YES NARROW NO RANGE LEVEL IN BOTH S!Gs LESS THAN (64%] 78% YES PRESSURE IN BOTH S/Gs LESS THAN 1085 PSIG Page 7 of 10 ***** *** NO YES NARROW NO RANGE LEVEL INBOTHSGs GREATER THAN[51%)32%
YES GOTO CSP-H.1 GOTO CSP-H.2 GOTO CSP-H.3 GOTO CSP-H.4 GOTO CSP-H.6 CSF SAT Calculation Numb e r 2004-006 10.0 Results and Conclus i ons 10.1 Dose Rates-RCS Activity R evision P ae 22 of 130 The initial dose rates (Rlhour) are summarized below. The RE-127 dose rates are plotted in Figure 12-1. RCS Activity:
300&#xb5;Ci/gm DE 1-131, 1 % fuel defects for non-iodine nuclides .... -... ~* . -*, *----***1 : 1/2RE-126 i 814.7 **----* -*----i------
*-* . : 1/2RE-12? .. 1-_?7~: 1. I i 1/2RE-128 -]-645.0 .. l : lRE-102 : 319.1 i ______ ,, __ .. ----,-**-* --***** . *. : 2RE-102 _______ L 553.3 _! 577 used based on Opeations feedback The dose rates have increased to about 34 times the Revision O values. RCS Activity:
1 % fuel defects for all nuclides 1/2RE-126 2.99E+Ol 1/2RE-127 2.12E+Ol 1/2RE-128 2.37E+Ol 1RE-102 l.17E+Ol 2RE-102 2.03E+Ol The dose rates have increased to about 1.6 times the Revis i on O values.
Calculation Number R ev i si on Pae 2004-006 23 of 130 10.2 Dose Rates-20%
Failed Fuel )/2RE-126 2.72E+04 , ~i/2RE-127
*--*---1.85E+04 ;1/2RE-128 f 2.20E+o4: 18,500 used based on Operations feedback ------* *--** --J, _________ .:J RE-127 Dose Rates are plotted in Figure 12-2. The dose rates have increased by 17% over the pre-EPU values. 10.3 Dose Rates-Technical Specification Activity l}./2RE-126_ ... i 1.42E+O~ ;1/2RE-127 j 1.10E+01! l 1/2RE:;2*s l 1.21E~oi l : 1RE-102 J}.}8~+QO*, 'iiE=iai * * . , 7_:&sect;7E+QQ.i 11 used based on Operations feedback RE-127 dose rates are plotted in Figure 12-3. The dose rates have increased by about 3.7 times the Revision O values. 10.4 Conclusions The estimated dose rates, for EPU conditions, have increased over those previously calculated for pre-EPU (Revision
: 0) conditions.
Comparison of the iodine Curies, provided in Sections 9.1 and 9.2, shows that 300 &#xb5;Ci/gm of DE I-131 (based on AST dose conversion factors) results in significantly greater activity than the I% defect level. Comparing the estimated dose rates from Section s 9. I and 9 .2 shows that 300 &#xb5;Ci/gm of iodine is a significant contributor to the dose rate. 11.0 Tables This section is empty.
Failed Fuel Monitor (RE-109) Reading/ Fuel Damage Correlation Results and
 
== Conclusions:==
* Calculation Revision 2004-0019 0 The results were calculated using a source term corresponding to a reactor thermal power operating level of 1683 MWth. The results are applicable generally to other thermal power operating levels because core activity and photon intensity are directly proportional to the reactor thermal operating power level. The failed fuel monitor reading at time equal _to O that would correspond to 300 &#xb5;Ci/g 1-131 DE is approximately 4500 mR/hr and the failed fuel monitor reading at time equal to O that would correspond to 50 &#xb5;Ci/g I-!'31 DE is approximately 750 mR/hr. The calculated RCS total radionuclide activity and fuel clad failure percentage correlations for use with the failed fuel monitor (RE-109) reading are graphically displayed in Figure 1 below. To determine fuel clad failure percentage, total primary system activity , or total 1-131 DE activity, multiply the respective conversion factor by the failed fuel monitor reading. Example: An event occurs causing fuel failure. The failed fuel monitor reading 12 hours after fuel failure occurs is 2000 mR/hour. The conversion factor for fuel clad failure from Figure 1 is approximately 7E-04 % clad failure per mR/hr. The estimated fuel clad failure is 7E-04 x 2000 or 1.4 %. These factors are acceptable for use when the sample system is not isolated , i.e., a containment isolation signal has not occurred. These factors can be used after sample system isolation has occurred, however, the values w i ll only reflect the condition of the primary system at the time of isolation.
* Figure 1 RE-109 Reading/Damage Relationship
' .. 1.00E+01 0 .... 0 1.00E+OO ' ... --a:s --u. i 1.00E-01 . T I C: 0 UJ 1.00E-02 I .. a, > 1.00E-03 I ' C: 0 0 1.00E-04 .. -I I I I I 0 4 8 12 . 16 20 24 Time From Fuel Clad Failure (hours) ...,._ 0/o clad failure per mR/hr -Total Act. (uCi/g per mR/hr) -+-1131 DE (uCi/gm per mR/hr) PBF-1608 Re visi on 7 08/05/04 Pag e 13 R c fo rc occ: NP 7 .2.4 POINT BEACH NUCLEAR PLANT CRITICAL SAFETY PROCEDURE CSP-ST.0 UNIT 1 SAFETY RELATED Revision 9 CRITICAL SAFETY FUNCTION STATUS TREES Page 8 of 10 TEI.IPERATURES NO IN BOTH COLD LEGS GREATER THAN 230&deg;F YES TEMPERA TURES IN BOTH COLDlEGS GREATER THAN 3 1 0'1' TEMPERATURE OECREASEIN NO BOTH COLO LEGS LESS THAN 100=F INTHE LAST YES 60MINUTES RCS PR ESSURE LESS THAN 415 PS I G TEMPERAT URES NO IN BOTH COLO LEGS GREATER THAN 2tS*f YES Q FIGURE 4 ST-4 INTEGRITY GOTO CSP-P.1 ----<, GOTO ... -CSP-P.1 NO GOTO CSP-P.2 YES -TEMPERATURES NO IN BOTH COLOU;GS GREATER THAN 34CYF YES CSF SAT GOTO CSP*P.1 TEl,IPEAA TURES NO IN BOTH COLO LEGS GREATER THAN 310'F Y ES NO GOTO CSP-P.2 YES Q 0 CSF SAT O CSF l SAT 
: 7. 8. Uplift due to buoyant forces External pressure load Containment System Structure FSAR Section 5.1 The critical loading condition is that caused by the maximum credible accident resulting from severance of a reactor coolant pipe coincident with the maximum hypothetical earthquake.
Loss of Coolant Accident L oad The design pressure and temperature of the containment is in excess of the peak pressure and temperature occurring as the result of the complete blowdown of the reactor coolant through any mpture of the reactor coolant system up to and including the hypothetical severance of a reactor coolant pipe. The suppmts for the reactor coolant system are designed to withstand the blowdown forces associated with the severance of the reactor coolant piping so that the coincidental rupture of the steam system is not considered credible.
Transients resulting from the loss of coolant accident and other lesser accidents are presented in Chapter 14 and serve as the basis for a containment design pressure of 60 psig. The design pressure is not exceeded during any subsequent long term pressure transient caused by the combined effects of such heat sources as residual heat and metal-water reactions.
These effects are overcome by the combination of emergency powered engineered safeguards and structural heat sinks. The temperature gradient through the wall during the loss of coolant accident is shown in Figure 5.1-6. The variation of temperature with time and the expansion of the liner plate are considered in designing for the thermal stresses associated with the loss of coolant accident load. Structure Dead Load Dead load consists of the weight of the concrete wall , dome, base slab, and any internal concrete.
Weights used for dead load calculations are as follows: 1. Concrete 2. Steel Reinforcing
& Prestressing Steel 3. Steel Lining UFSAR2017 143 lb/ft 3 489 lb/ft 3 using nominal cross sectional areas of reinforcing as defined in ASTM for bar sizes and nominal cross sectional areas of prestressing tendons. 489 lb/ft 3 using nominal cross sectional area of lining. Page 5.1-32 of 109 POINT BEACH NUCLEAR PLANT SETPOINT DOCUMENT EMERGENCY OPERA TING PROCEDURE (EOP) SETPOINTS PNBP ERG STPT STPT PARAMETER DESCRIPTION L.26 Flow (Charging)
Maximum c har ging flow L.27 Flow(AFW)
Minimum indicated AFW flow to mitigate LONF/LOAC event , including normal uncertainties L.28 S.09 Flow (RHR) RHR pump flowrate required to sweep air from the RCS hot leg piping, inc ludin g allowance for normal channel accuracies. L.29 S.11 F low (Charging)
Charging flowrate requirement that discriminates between an RCS leak and a sma ll break Loss of Coolant Accident minus the flow rate of the highest capac ity letdown orifice, not to exceed maximum VCT makeup capacity. M.1 Containment Containment High-High pressure SLI setpoint (Pressure)
.2 T.02 Co nt a inm e n t Co nt a inm e n t w ess ur e se t o int for ~ay ac tu at i o n Pr ess ur e M.3 SPARE M.4 T.03 Containment Containment design pressure (Radiation) M.5 T.07 Conta inm ent Radiation level alarm setpoint for post-accident containment radiation monitor (Radiation)
M.6 SPARE M.7 T.06 Conta inm ent Sump B water level just below design flood level (S ump) M.8 SPARE Page 17 of28 STPT 25.1 Revision 14 SETPOINT U nit 1 U nit2 140 gpm 140 gpm 315 gpm 315 gpm 1550 gpm 1550 gpm 30 gpm 30 gpm 15 psig 15 psi g 2 5 sig 25 _Qs i g 60 psig 60 psi g 10 R/hr 10 R/hr 86 in 86 in CALC REFERE CE 2010-0022 2010-0022 2010-0022 2010-0022 2010-0023 20 1 0-0023 2010-0023 2010-0023 2010-0023 2010-0023 2010-0023 2010-0023 POINT BEACH NUCLEAR PLANT PROCEDURES MANUAL EXTERNAL EVENTS PROGRAM DISCUSSION/OVERVIEW:
APPENDIX I SEISMIC Pagel of 4 NP 7.2.29 Revision 5 Earthquakes are defined as a vibration of the eaith's surface that occurs after a release of energy in the earth's crust. Because the earth's crust is made up of numerous segments or "plates" that are constantly moving slowly, vibrations can occur and result in small or large earthquakes.
Most earthquakes are quite small and not readily felt. Larger and more violent earthquakes are those that occur in a release of energy as the plates slide past or collide into one another. PURPOSE: This document is intended to identify those seismic protection features in-place at the site, and to identify the implementing document(s) that are used to check their function and/or ensure they are capable of continuing to perform their function.
CLB FEATURES:
Horizontal ground acceleration at the site of 0.06g combined with a vertical acceleration of 0.04g are used for the Operating Basis Ea1thguake (OBE). These accelerations are considered as acting simultaneously.
OBE is selected to be typical of the largest probable grou.nd motion based on the site seismic history. Components and systems that are essential for continued operation without undue risk to the health and safety of the public are designed to remain functional due to OBE seismic effects. The hypothetical earthquake or Safe Shutdown Ea1thquake (SSE) is selected to be the largest potential ground motion at the site based on seismic and geological factors and their unceitainties.
SSE is twice the magnitude of the OBE. The seismic design for the SSE is intended to provide a margin in design that assures the integrity of the reactor coolant pressure boundary, the capability to shutdown the reactor and maintain it in a safe shutdown condition, the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the exposures of 10 CFR 50.67. EVENT RESPONSE:
Seismk Detector Annunciator COl A 2-6 and PPCS Seismic Alarm will go in alarm if O.Olg horizontal or ve1tical acceleration is recorded.
PPCS Seismic Alarm will alarm if 0.06g horizontal acceleration or a 0.04g vertical acceleration is recorded.
Either alaim will prompt entry into AOP-28, Seismic Events, as will the event being "felt" by the operations staff, or reported to have occurred by a reliable source. A plant walkdown to determine equipment damage is performed while the data from the seismic recorders is being retrieved and analyzed.
If it is determined by the seismic recorder analysis that an Operating Basis Eatthquake (OBE) limit has been exceeded, then a focused inspection by Engineering will be performed using PBF-7044, Focused Inspection.
Page 45 of 52 INFORMATION USE Matrix of Table H-1 Fire Areas below was developed by review of AOP-40A, "Control Room Abandonment Due To Fire" and the PBNP Shutdown Safety Plan. Listing was then verified against the Vital Areas of the PBNP Security Plan to identify any other potential areas containing safety-related equipment.
AOP-40A Shutdown Safety Plan Security Vital Areas G05 Building X X 13.8kV building X X Cable Spreading Room X X X Vital Switchgear Room X X X AFWRoom X X X G01/G02 Rooms X X X EDG building X X X P AB (all elevations)
X X X Service Water Pump Rooms X X Containment X X Control Room X X X Facade 85' X X (by default behind P AB VA door) The resulting site-specific list of plant rooms or areas for PBNP was captured in Table H-1 below: Table H-1 Areas Control Room Containment PAB G05 building 13.8kV Building Cable Spreading Room Vital Switchgear Room AFW Pump Room G-01/02 Rooms EDG Building Service Water Pump Rooms Fa9ade 85' Local Operations for Normal Operation/Shutdown/Cooldown Step-by-step analysis was performed on the following PBNP procedures to determine those "rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, coo/down and shutdown":
* OP 3A: Power operation to hot standby
* OP 38: Reactor Shutdown
* OP 3C: Hot Standby to Cold Shutdown
* OP ?A: Placing RHR System in operation
* OP 50 Part 4: Degassing the RCS using the PZR and Letdown Gas Stripper Analysis did not include rooms or areas in which actions of a contingent or emergency nature would be performed. (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations).
Analysis also did not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).
Backup to CR or Vital to Procedure Procedure:
Step: Mode: Action: automatic plant Location of vital action: action? ops OP3A 5.2.3 1 Start MFP seal water pumps y N OP3A 5.4.4 1 SW overboard alignment N N OP3A 5.5.3 1 shut blast damper N N OP3A 5.9.3 1 Open MSR purge valves N N OP3A 5.11.1 1 Bypass LP Feed heater coolers N N OP3A 5.11.5 1 shut mov-1 and 2 y N OP3A 5.13.5 1 start turbine bearing lift oil y N pumps OP3A 5.20.4 1 lube oil cooldown n N ' ',, ' OP 38 none n/a n/a n/a n/a V Backup to CR or Vital to Procedure:
Procedure Mode: Action: automatic plant Location of vital action: Step: action? ops OP 3C 5.1.1 3/4/5 Put the flex pumps in 8' fan N N room OP 3C 5.1.7 3/4/5 Sample blender output N y VCT Area OP 3C 5.1.8 3/4/5 degas the RCS N y See OP SD P4 below OP 3C 5.6 3/4/5 N2 to the VCT N y VCT A rea OP 3C 5.12 4/5 Isolate accumulators N y C-59 area OP3C 5.13 4 align fle x accumulator N N OP3C 5.16 4 Isolate SI pump y N backup act i ons to CR only OP3C 5.23.6 5 Align containment purge N N OP 3C 5.23.7 4 align flex N2 N N Restraint fo r entry into Mode 5 only, plant is stable and can remain in this condition until area accessible OP 3C attachment A 3/4/5 borate to refuel i ng concentration y N backup to CR only Stop at step 5.24 because plant is now in mode 5.
Backup to CR or Vital to Procedure:
Procedure Mode: Action: automatic plant Location of vital action: Step: action? ops OP7A 5.l.3a 4 adjust CC cooling to RHR N y Pipeway 2/3, 8' elev. OP7A 5.1.4 4 caution tag 851871s N y C-59 area OP7A 5.1.5 4 align RHR suction N y C-59 area , -Sft OP7A 5.1.6 4 shut RH-716 N N OP 7A 5.2.16 4, 5 CCW temp control N y CCW HX Room OP SD P4 5.2 3 Initiate primary degas N y primary sample room OP SD P4 5.3.3 3 tag shut CV-261c N y VCT area OP SD P4 5.3.5 3 isolate H2 to vet N y VCT area Resulting Tables used in EALs RA3 and HAS are shown below: Table R-2 SAFE OPS, S/D, C/D AREAS .. ~rea/Bu .. jlding* __ . _MODE .. , ... -* Ul VCT Area 3/4/5 U2 VCT Area 3 /4/ 5 Ul Primary Sample area 3 U2 Primary Sample area 3 CCW HX Room 4/5 C-59 area 3/4/5 Pipeway 2, 8 ft. Elev. 3/4 Pipeway 3, 8 ft. Elev. 3/4 1/2B-32 MCC Area 4 Table H-2 SAFE OPS, S/D, C/D AREAS ....... Area/Building.
MODE Ul VCT Area 3/4/5 U2 VCT Area 3 /4/ 5 Ul Primary Sample area 3 U2 Primary Sample area 3 CCW HX Room 4/5 C-59 area 3/4/5 Pipeway 2, 8 ft. Elev. 3/4 Pipeway 3, 8 ft. Elev. 3/4 1/2B32 MCC Area 4 This validation document (V29) is not used Fa i led Fuel Monitor (RE-109) Reading/ Fuel Damage Correlation Results and
 
== Conclusions:==
* Calculation Revision 2004-0019 0 The results were calculated using a source tenn corresponding to a reactor thennal power operating level of 1683 MWth. The results are applicable generally to other thennal power operating levels because core activity and photon intensity are directly proportional to the reactor thermal operating power level. The failed fuel monitor reading at time equal .to O that would correspond to 300 &#xb5;Ci/g I-131 DE is approximately 4500 mR/hr and the failed fuel monitor reading at time equal to O that would corresQ_Ond to 50 &#xb5;Ci/g 1-131 DE is approximately 750 mR/hr. The calculated RCS total radionuclide activity and fuel clad failure percentage correlations for use with the failed fuel monitor (RE-109) reading are graphically displayed in Figure 1 below. To determine fuel clad failure percentage, total primary system activity, or total I-131 DE activity, multiply the respective conversion factor by the failed fuel monitor reading. Example: An event occurs causing fuel failure. The failed fuel monitor reading 12 hours after fuel failure occurs is 2000 mR/ho~r. The conversion factor for fuel clad failure from Figure 1 i s approximately 7E-04 % clad failure per mR/hr. The estimated fuel clad failure is 7E-04 x 2000 or 1.4 %. These factors are acceptable for use when the sample system is not isolated, i.e., a containment isolation sign a l has not occurred.
These factors can be used after sample system isolation has occurred , however, the values will only reflect the condition of the primary system at the time of isolation.
Figure 1 RE-109 Reading/Damage Relationship
' .. 1.00E+01 0 ... u 1.00E+OO ..:... -a:s -IL 1.00E-01 --I I C: 0 *-en 1.00E-02 .. Q) > 1.00E-03 I C: I 0 0 1.00E-04 -I I I 0 4 8 12 16 20 24 Time From Fuel Clad Failure (hours) -A-0/o clad failure per mR/hr -Total Act. (uCi/g per mR/hr) -+-1131 DE (uCi/gm per mR/hr) PBF-1 6 08 Re visi on 7 08/0S/04 P age 13 R c fcr c occ: NP 7 .2.4 
 
===3.4 REACTOR===
COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity RCS Specific Activity 3.4.16 LCO 3.4.16 RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT Xe-133 specific activity shall be within limits: APPLICABILITY:
MODES 1 , 2 , 3 , and 4. ACTIONS CONDITION A. DOSE EQUIVALENT 1-131 not within limit. B. DOSE EQUIVALENT Xe-133 not within limit. Point Beach REQUIRED ACTION COMPLETION TIME ----------
--------Note-
-------------
---LCO 3.0.4.c is applicable. -----------------------------------------
A.1 Verify DOSE Once per 4 hours EQUIVALENT 1-131 S50 &#xb5;Ci/gm. AND A.2 Restore DOSE 48 liours EQUIVALENT 1-131 to within limit. ------------------Note-----------------
48 hours LCO 3.0.4.c is applicable. ------------------------------------
-----B.1 Restore DOSE EQUIVALENT Xe-133 to within limit. (continued) 3.4.16-1 Unit 1 -Amendment No. 233 Unit 2 -Amendment No. 238 RCS Specific Activity 3.4.16 ACTIONS (continued)
-~---~---------------------
CONDITION C. Required Action and C.1 associated Completion Time of Condition A or B AND not met. OR DOSE EQUIVALENT 1-131 >50 &#xb5;Ci/gm. C.2 REQUIRED ACTION Be in MODE 3. Be in MODE 5. SURVEILLANCE REQUIREMENTS SURVEILLANCE COMPLETION TIME 6 hours 36 hours FREQUENCY SR 3.4.16.1 -----------
----------------NOTE----------------------
----In accordance SR 3.4.16.2 I Point Beach Only required to be performed in MODE 1. with the Verify reactor coolant DOSE EQUIVALENT Xe-133 Specific Activity~
300 &#xb5;Ci/gm. ----------------------
-----NOTE--------------------------
0 n I y required to be performed in MODE 1. Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity:,:; 0.5 &#xb5;Ci/gm. Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program AND Between 2 and 6 hours after a THERMAL POWER change of~ 15% RTP within a 1 hour period 3.4.16-2 Unit 1 -Amendment No. 253 Unit 2 -Amendment No. 257 Excerpt from NEI 99-01 Rev6: EAL SU4 (SU5 BNP) Deve l oper Notes: EAL #1 (unidentified leakage)-For the site-specific leak rate value, enter the higher of 10 g m or the value s ecified in the site's Technical S ecifications for this t)'. e of leakage. PBNP Tech Specs Jim it for unidentified leakage is 1 gpm; therefore 10 gpm is entered as specified above EAL #2 (identified leakage)-For the site-specific leak rate value , enter the higher of 25 g m or the value s ecified in the site's Technical S ecifications for this ty___pe of leakage. PBNP Tech Specs limit for identified leakage is 10 gpm; therefore 25 gpm is entered as specified above 
 
===3.4 REACTOR===
COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE RCS Operational LEAKAGE 3.4.13 LCO 3.4.13 RCS operational LEAKAGE shall be limited to: a. No pressure boundary LEAKAGE; b. 1 gpm unidentified LEAKAGE: c. 10 9pm Identified LEAKAGE; and d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG). APPLICABILITY
: MODES 1, 2 , 3 , and 4. ACTIONS CONDITION A. RCS operational LEAKAGE not within limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE. B. Required Action and associated Completion Time of Condition A not met. OR Pressure boundary LEAKAGE exists. OR -Primary to secondary LEAKAGE not with i n limit. Point Beach A.1 B.1 AND B.2 REQUIRED ACTION COMPLETION llME Reduce LEAKAGE to 4 hours within l i mits. Be in MODE 3. 6 hours Be in MODE 5. 36 hours 3.4.13-1 Unit 1 -Amendment No. 223 Unit 2 -Amendment No. 229 POINT BEACH NUCLEAR PLANT SETPOINT DOCUMENT EMERGENCY OPERA TING PROCEDURE (EOP) SETPOINTS PNBP ERG STPT STPT V.39 V.40 V.41 V.42 .4 3 V.44 V.45 V.46 X.04 W.l W.2 Z.1 Z.2 G.03 Z.3 PARAMETER Shutdown Margin RCP Seal dP Main Steam Pressure EDG Frequency
-0 l /l)-02 oltage No. ofSGs RHR Pump Discharge pressure us RCS Temperature (Core Exit) RCS Temperature (Core Exit) PZR Level DESCRIPTION Shutdown margin for RCS temperature at 70&deg;F Minimum RCP seal differential pressure for normal RCP seal cooling flow Minimum pressure differential between intact and ruptured SGs Maximum allowable emergency diesel generator frequency
-01 /lJ-02 minimum battery bus voltage SPARE SPARE Number of steam generators necessary to maintain RCS pressure low enough to gravity feed. Minimum RHR pump discharge pressure to maintain pump flow less than 2200 gpm to ensure pump does not reach runout conditions SPARE Core exit temperature indicative of inadequate core cooling (ERG value given) Temperature correspond in g to 670&deg;F plus normal channel accuracy and post-accident transmitter errors or 700&deg;F , whichever is greater Pressurizer level range to ensure adequate inventory to accommodate void growth Page 26 of28 STPT 25.1 Revision 14 SETPOINT Unit 1 Unit 2 1% ~k/k 1% ~k/k 20 in w.c. 20 in w.c. 250 psid 250 psid 60.3 Hz 60.3 Hz 115 ac 115 ac 120 psig 120 psig l 200&deg;F 1200&deg;F 700&deg;F 700&deg;F 20% to 30% 20% to 30% CALC REFERENCE 2010-0028 2010-0028 2010-0028 2010-0028 2010-0028 2010-0028
 
2010-0028 2010-0028 2010-0028 2010-0028 2010-0017 2010-0017 2010-0016 A.I STATION BLACKOUT (SBO) A.I.I STATION BLACKOUT OVERVIEW Station Blackout FSAR Appendix A. I Station Blackout is defined as the complete loss of alternating current electric power to the essential and nonessential switchgear buses in a nuclear power plant (i.e., loss of offsite electric power system concurrent with a turbine trip and the unavailability of the onsite emergency AC power system). A Station Blackout does not involve the loss of available AC power to buses fed by station batteries through inverters.
The event is considered to be terminated upon the restoration of power to the essential switchgear buses from any source, including the alternate AC source which has been qualified as an acceptable coping mechanism.
A concurrent single failure or design basis accident need not be assumed during a station blackout event (Reference 2 and Reference 18). The requirements for Station Blackout are established in 10 CPR 50.63 (Reference 1), which was formally issued in 1988. Guidance for compliance with the regulatory requirements is presented in NUMARC 87-00, Revision O (Reference
: 2) and Regulatory Guide 1.155 (Reference 3). The NRC has not endorsed Revision 1 to NUMARC 87-00, but has accepted specific supplements to NUMARC 87-00 Rev. 0, as described in Appendix K ofNUMARC 87-00 Rev. 1 (Reference 2). The station blackout regulation requires determination of the coping duration category based on criteria provided in Reference 2 and Reference
: 3. The "required coping duration" is defined as the time between the onset of station blackout and the restoration of off-site AC power to safe shutdown buses. "Coping duration category" is a quantification of the relative risk of a particular facility to the occurrence of a station blackout (loss of all onsite and offsite AC power). The determination of the required coping duration category is based on several factors, such as the plant design and the probability of severe weather conditions in the area. Once the required coping duration category has been established, the design approach to coping with the station blackout event is demonstrated.
This design approach may choose to take credit for either an available alternate AC power source or opt for an AC power-independent design. The plant systems must have the necessary capacity and capability to ensure the core is cooled and containment integrity is maintained for the required station blackout coping duration.
The coping duration categories are 2, 4, 8, or 16 hours, as determined from Table 3-8 of Reference
: 2. The intent of the regulation is for all domestic nuclear plant sites to fall in either the 2-hour or the 4-hour coping duration category, and then select either the "Alternate AC" or "AC-Independent" coping methodology for their specific plant. The NRC bases for coping duration category objectives are described in Section 2.3.2 of Reference
: 2. The major contributor to overall station blackout risk is the likelihood of losing off-site power and the duration of power unavailability.
The stated objective of the NRC is to reduce the core damage frequency due to station blackout to approximately 1 o*5 per year for the average site. This objective is accomplished by requiring either a four hour coping capability or use of an Alternate AC (AAC) source. PBNP's original response to the SBO rule concluded that the required coping duration category was 8 hours and used the Gas Turbine Generator (GTG) G-05 as the sole Alternate AC (ACC) source to power the safe shutdown loads of both blacked out units. Because the GTG cannot be UFSAR2013 Page A.I-I of 11 Station Blackout FSAR Appendix A.1 shown to be available within 10 minutes of the onset of station blackout, a one hour coping assessment was performed as required by Section 7.1.2 of Reference
: 2. (Reference 4 and Reference
: 8) The coping duration category was subsequently revised to 4 hours based on a change in the extremely severe weather (ESW) group classification as discussed in Section A.1.2. With the addition of the G-03 and G-04 EDGs, the SBO minimum redundancy requirements of emergency AC (EAC) power supplies for normal safe shutdown of both units is exceeded and utilization of an EDG as an AAC source is allowed. By definition, a unit with an available EAC power supply is not blacked out. However, any EDG credited as an AAC source must be capable of handling the safe shutdown loads in both the blacked out and non-blacked out units (Reference 2). The PBNP EDGs meet this requirement.
Therefore, the present coping methodology utilizes the Gas Turbine Generator (GTG) G-05 or an Emergency Diesel Generator (EDG) from the non-blacked out unit as Alternate AC (ACC) sources. An EDG will start, accelerate to rated frequency and voltage, and can be connected to an EAC bus in either unit within ten minutes of SBO initiation.
The GfG will be manually started, accelerate to rated frequency and voltage, and be available to power the safe shutdown loads within one hour of SBO initiation (Reference 6, Reference 15). Since PBNP continues to use the GfG as one of the ACC sources, and it cannot be shown to be available within 10 minutes, the one hour coping assessment has been retained and is described in Section A.1.3. A.1.2 STATION BLACKOUT COPING DURATION CATEGORY DETERMINATION The potential for long duration loss of off-site power (LOOP) events can have a significant impact on station blackout risk and required coping duration.
Long duration LOOP events are typically associated with grid failures due to severe weather conditions or unique transmission system features.
Shorter duration LOOP events tend to be associated with plant specific switchyard features.
Per Reference 1, the required coping duration shall be based on the following factors: 1. The redundancy of the emergency standby power system 2. The reliability of each of the emergency power sources 3. The expected frequency of a loss of off site power 4. The probable time required to restore offsite power Offsite Power Design Characteristic Group The regulatory guidance (Reference 2, Tables 3-5a and 3-6a; Reference 3, Table 4) has established three basic groups (Pl, P2, and P3) for categorizing the design of the preferred offsite power system. A category of P3 is assigned to those plants with a frequency of grid-related loss of off site power events greater than once in 20 site-years, which is limited to St. Lucie, Turkey Point and Indian Point (Reference 2). Since PBNP is not included among the three noted plant sites, further evaluation of several factors is necessary to establish the Offsite Power Design Characteristic Group. The applicable group is defined based on combinations of the following three factors:
* extremely severe weather
* severe weather
* offsite power system independence UFSAR2013 Page A.1-2 of 11 Extremely Severe Weather (ESW Group) Station Blackout FSAR Appendix A. I The estimated frequency of loss of offsite power due to extremely severe weather is determined by the annual expectation of storms at the site with wind velocities equal to or greater than 125 mph. These events are normally associated with the occurrence of hurricanes where high windspeeds may cause widespread transmission system unavailability for extended periods. Since electrical distribution systems are not designed for such conditions, it is assumed the occurrence of such windspeeds will directly result in the loss of offsite power. The estimated frequency may be determined based on either site-specific data or on data from local weather stations.
Table 3-2 of Reference 2 summarizes site-specific National Oceanic Atmospheric Administration (NOAA) data for the estimated frequency of occurrence of extremely severe weather. As published in this table, PBNP has an event frequency of 0.0036, and therefore was categorized in ESW Group 4 (Reference 4). Subsequent review determined the NOAA data for extremely severe weather was overly conservative for PBNP, and that an ESW event frequency supporting an ESW Group 2 category was justified (Reference 5). This departure from the NUMARC 87-00 criteria was reviewed and approved by the NRC (Reference 6). Severe Weather (SW Group) Table 6 of Reference 3 and Part 3.2.1.C of Reference 2 define the severe weather factor based on the frequency of a loss of off site power due to severe weather. The severe weather considered includes snow, tornadoes, high winds, and storms with salt spray. These are related by the equation:
frequency=
1.3 x 10-4 x h1 +bx h2 + 0.012 x h3 + c x h4 The variables in this equation are defined for PBNP in Reference 2, Section 3.2.1.C: h 1 = annual expectation of snowfall for site, in inches; this is 42.0 inches for PBNP h 2 =annual expectation of tornadoes with windspeeds greater than or equal to 113 miles per hour, in events per square mile; this is 0.000035 for PBNP h 3 =annual expectation of storms with wind velocities between 75 and 124 mph; this is 0.1 for PBNP h 4 =annual expectation of storms with significant salt spray for the site; this is 0.0 for PBNP. b =72.3; the PBNP offsite power system design connects four 345 kV transmission circuits to the plant switchyard via a single right-of-way.
c =O; the PBNP site is not considered vulnerable to the effects of salt spray. These factors, when combined in the severe weather frequency equation, yield an estimated frequency of loss of off site power due to severe weather of 0.0092. This places PBNP in SW Group 2. UFSAR2013 Page A.1-3 of 11 Independence of the Off site Power System (I Group) Station Blackout FSAR Appendix A. I Reference 3, Table 5, defines the offsite power system independence factor, and Reference 2 Section 3.2.1.D simplifies the determination:
If: (a) all offsite power sources are connected to the safe shutdown buses through one switchyard or through multiple electrically connected switchyards, and (bl) the normal power source is from the main generator and there are no automatic and one or more manual transfers of all safe shutdown buses to the preferred or alternate offsite power sources, or (b2) there is one automatic and no manual transfers of the safe shutdown buses to one preferred or one alternate offsite power source, the site falls in the I-3 group. Otherwise, the site is assigned to the I-1/2 group. The I-1/2 group is characterized by features associated with greater independence and redundancy of sources, and a more desirable transfer scheme. I-3 sites have simpler, less desirable offsite power systems and switchyard capabilities.
Condition a: The PBNP offsite power system consists of four (4) 345 kV transmission circuits, connected via a single right-of-way, to a single switchyard which serves both PBNP units. On this basis, the answer to Condition A is considered to be "YES" for the PBNP site. Condition b 1 and b2: The PBNP auxiliary power distribution system provides offsite power connections to the safety-buses of each unit via the high voltage station auxiliary transformers and the low voltage station auxiliary transformers.
This normal supply of power to the safety-related buses is derived from offsite power sources. Upon loss of the preferred offsite power source to the safety-related buses of one unit, the buses will be powered from the preferred power source of the other unit. On this basis, the answers to both Condition b(l) and b(2) are considered to be "NO", and the PBNP site is classified in the I-1/2 Group. Offsite AC Power Design Characteristic Group Determination The combination of the ESW, SW and I factors results in an Offsite Power Design Characteristic Group of Pl for PBNP, based on Reference 3, Table 4. Emergency AC Power Configuration Group Regulatory guidance defines four Emergency AC (EAC) Power Configuration groups (A, B, C, and D) based on the availability and redundancy of the emergency power supplies.
Reference 2 Section 3.2.2 clarifies the EAC groups, basing it on the number ofEAC power supplies required to handle the safe shutdown loads and on the number of additional EAC power supplies available.
The PBNP EAC power configuration group is C, based on the following:
PBNP is a two-unit site with four shared Emergency Diesel Generators (EDGs) and one gas turbine generator (GTG). The two Train A EDGs are identical components with a 2000 hour rated output of2850 kW at 4.16 kV. The two Train B EDGs are identical components with a 2000 hour rated output of2848 kW at 4.16 kV. All four EDGs are available to support the safe shutdown equipment of either PBNP unit, and a single EDG can supply adequate power to the safe shutdown loads in both units. The GTG has a rating of 23 .10 MVA at an output voltage of 13.8 kV, and can supply adequate power to the safe shutdown loads in both units. UFSAR2013 Page A.1-4 of 11 Station Blackout FSAR Appendix A. l Therefore, because only one EDG is necessary to operate safe shutdown equipment for both units following a loss of offsite power, the EAC power configuration group at PBNP is "C", as a 1 out of2 EDG, dedicated, or 1 out of3 EDGs, shared configuration per Table 3-7 of Reference
: 2. Additionally, the PBNP SBO licensing basis permits the use of either the GfG or an EDG as the AAC source. Target Standby Diesel Generator Reliability The reliability of the EAC power sources has a key role in the quantification ofrisk due to SBO. A target value for reliability was therefore made a factor in establishing the required SBO coping duration.
The EDG target reliability was selected to be 0.975 based on the original EAC configuration group determination of"D" (i.e., prior to the installation of G-03 and G-04) and the reliability data that existed at the time of the initial SBO evaluation. (Reference
: 4) These reliability computations utilized the NRC-recommended methodology of EPRI Report NSAC-108 (Reference
: 7) Because PBNP offsite power design group is Pl, and EAC configuration is C, the target EDG reliability value may be 0.950 or 0.975 per Table 2 of Reference
: 3. PBNP has retained the reliability target value of 0.975 (Reference 15). PBNP has implemented an EDG reliability program which is based on the methodology of EPRI Report NSAC-108 and conforms to the guidance ofRG 1.155, Position 1.2 (Reference 8 and Reference 15). Coping Duration Category Determination Summary The previous determinations are summarized below: Offsite AC Power Design Characteristic Group =Pl Emergency AC Power Configuration Group =C Target Standby Diesel Generator Reliability
=0.975 ,Irt accordance
~ith Table 3-8 ofRefe;ence 2 a~d Table 2 of Reference 3: the group, determ.inations list_ed above resuJt in a coping duration categs,ry for PBNP of _four houi:s~ A.1.3 STATION BLACKOUT COPING ANALYSES Condensate Inventory for Decay Heat Removal This analysis ensures that PBNP has sufficient condensate inventory to support the decay heat removal function for the SBO event duration.
Section 7.2.1 of Reference 2 provides a simplified calculation approach to determine the required condensate volume. This analysis is satisfied by demonstrating that Technical Specification volume requirements envelop the volume estimated by the Reference 2 methodology.
At a core power of 1800 MWt, 14,000 gallons of condensate water are required for the one hour SBO event duration based on the methodology of Reference
: 2. However in order to maintain the same margin set by the NRC in Reference 8 for subsequent switchover to the long-term AFW water supply, the minimum CST usable volume is set at 15,410 gallons. This volume is bounded by the Technical Specification CST volume requirements which includes additional margin to UFSAR2013 Page A.1-5 of 11 1-PT-FP-001 UNIT 1 CONTAINMENT FIRE DETECTOR TEST (REFUELING)
DOCUMENT TYPE: Technical CLASSIFICATION:
Safety Related REVISION:
3 REVIEWER:
NIA APPROVAL AUTHORITY:
Department Manager PROCEDURE OWNER (title): Group Head OWNER GROUP: Operations Verified Current Copy: ____________
_ Signature Date Time List pages used for Partial Performance Controlling Work Document Numbers Completed Procedure Review: Shift Supervision (Print) Shift Supervision (Sign) Date Time POINT BEACH NUCLEAR PLANT PERIODIC TEST FIRE PROTECTION PROCEDURE UNIT 1 CONTAINMENT FIRE DETECTOR TEST (REFUELING)
TABLE OF CONTENTS SECTION TITLE 1-PT-FP-001 SAFETY RELATED Revision 3 PAGE 1.0 PURPOSE .......................................................................................................................
3 2.0 PREREQUISITES
..........................................................................................................
3 3.0 PRECAUTIONS AND LIMITATIONS
.........................................................................
3 4.0 INITIAL CONDITIONS
................................................................................................
4 5.0 PROCEDURE
.................................................................................................................
5 6.0 ACCEPTANCE CRJTERIA ...........................................................................................
6
 
==7.0 REFERENCES==
 
...............................................................................................................
6 8.0 BASES ............................................................................................................................
6 ATTACHMENT A DETECTOR RESPONSE .....................................................................................
8 ATTACHMENT B DETECTOR LAYOUT .........................................................................................
9 Page 2 of 11 REFERENCE USE POINT BEACH NUCLEAR PLANT PERIODIC TEST 1-PT-FP-001 SAFETY RELATED Revision 3 FIRE PROTECTION PROCEDURE UNIT I CONTAINMENT FIRE DETECTOR TEST (REFUELING)
CFP -----1.0 PURPOSE 1.1 This test is to verify the integrity and standby operability of the Unit I containment smoke detection system. 1.2 This test is required by NFPA 72, National Fire Alarm Code and NEIL Member's Manual. 2.0 PREREQUISITES NOTE: The steps in the prerequisites section may be performed in any sequence.
2.1 The fire detection system including the fireworks stations are available for testing the detectors.
 
===2.2 Radiation===
 
Protection has been notified of the areas of containment that will be accessed and Radiation Work Permit has been obtained.
 
===2.3 Special===
Tools and Equipment
* Smoke Detector Test Pole -Solo 301-024 or equivalent (I&C M&TE Locker)
* Smoke Detector Tester -Stock 915-2593, 2 cans (OPS Flammable Locker)
* Extension Ladder
* Two radios 3.0 PRECAUTIONS AND LIMITATIONS
: 3. I Any testing performed greater than or equal to 6 feet above floor level requires the use of personnel safety equipment (e.g., safety belt, etc.). 3.2 Detectors that have alarmed will NOT alarm again after acknowledging and are considered Out-Of-Service till Reset. Alarms should be reset prior to exiting containment.
3.3 Any deficiencies should be immediately reported to the Shift Management for appropriate action in accordance with OM 3.27. Page 3 of 11 REFERENCE USE POINT BEACH NUCLEAR PLANT PERIODIC TEST FIRE PROTECTION PROCEDURE UNIT 1 CONTAINMENT FIRE DETECTOR TEST (REFUELING)
 
===4.0 INITIAL===
CONDITIONS NOTE: ALARA shall be considered at all times 4.1 Unit 1 is in any condition.
1-PT-FP-001 SAFETY RELATED Revision 3 4.2 Radiation Protection has been notified to provide support, ifrequired.
4.3 A pre-job brief has been performed.
4.4 This test is being done to satisfy: __ The normally scheduled callup. Task Sheet No. _____ _ NOTE: If this test is being performed to satisfy PMT or off-normal frequency requirements, Shift Management may NIA those portions of the procedure that are NOT applicable for the performance of the PMT. The use ofN/A is NOT acceptable for Initial Conditions, Precautions and Limitations, or procedure steps that pertain to the equipment requiring PMT, nor is it acceptable for restoration of equipment/components unless the component has been declared inoperable.
__ Post maintenance operability test Equipment ID _____ _ WO No(s). _____ _ Task Sheet No.(s) ______ _ __ Special test -no numbers Explain: ________________
_ 4.5 Permission to Perform Test INITIALS The conditions required by this test are consistent with required plant conditions, including equipment operability.
Permission is granted to perform this test. Shift Management Date* Time Page 4 of 11 REFERENCE USE POINT BEACH NUCLEAR PLANT PERIODIC TEST FIRE PROTECTION PROCEDURE UNIT 1 CONTAINMENT FIRE DETECTOR TEST (REFUELING) 1-PT-FP-001 SAFETY RELATED Revision 3 NOTE: The first detector smoked will require a longer smoke time since the detector will first go into the Alert Activation condition before the Alarm Activation.
NOTE: The smoker of the detector should NOT announce the detector to be tested. After the detector is smoked, the individual at the Fireworks station will caffout the detector in alarm and the smoker will confirm it. 5.0 PROCEDURE NOTE: Annunciator Window 1C20 A 1-2, FACP ALARM, will be received with the first detector tested. 5.1 Ensure Control Room has been informed of the fire alarms that will come in during test performance.
5.2 At the Fireworks station, go to the sitemap screen to begin the test for a given elevation.
NOTE: Attachment B may be referenced for detector location.
 
===5.3 Using===
the Smoke Pole, smoke a detector.
5.4 WHEN the detector goes into Alarm Activation status with the outlined area blinking in red,
* THEN perform the following:
 
====5.4.1 Control====
Room Operator Acknowledge Annunciator Window 1C20 A 1-2, FACPALARM.
NOTE: For each elevation, it is only necessary to drill down to the appropriate Containment elevation screen for the first detector tested. 5.4.2 On the Containment screen that shows the individual detectors, confirm detector in alarm. 5.4.3 Touch or click on "ACK" in the bottom right hand comer. 5.4.4 To confirm the detector, radio the detector smoker the detector that went into alarm. 5.4.5 On Attachment A, initial for the detector that went into alarm. Page 5 of 11 REFERENCE USE POINT BEACH NUCLEAR PLANT . PERIODIC TEST FIRE PROTECTION PROCEDURE 1-PT-FP-001 SAFETY RELATED Revision 3 UNIT 1 CONTAINMENT FIRE DETECTOR TEST (REFUELING)
NOTE: The preferred means of testing is to reset the alarms after all detectors have been tested. Resetting detectors will require the first detector after resetting to go through the Alert Activation condition slowing down the testing. 5.5 Repeat Steps 5.3 through 5.4.5 for a given elevation.
5.6 WHEN an elevation is complete, THEN repeat Steps 5.2 through 5.4.5. 5. 7 WHEN all the detectors have been tested, THEN ensure the detectors are reset on the control room Fireworks PC using the reset in the bottom left hand comer of the touch screen.
* 6.0 ACCEPTANCE CRJTERJA 6.1 All the detectors tested in Attachment A went into the alarm activation condition when smoked.
 
==7.0 REFERENCES==
 
7.1 NFPA 72, National Fire Alarm Code 7.2 NEIL Member's Manual 7 .3 OI 40A, Fire Alarm Control Panel and Fireworks PC Operation 7.4 FPPDD, NFP A 805 Fire Protection Program Design Document 7.5 OM 3.27, Control of Fire Protection
& NFPA 805 Equipment
 
===7.6 Drawings===
7.6.1 EST FPE-001 7.6.2 EST FPE-002 7.6.3 EST FPE-003 8.0 BASES None Page 6 of 11 REFERENCE USE POINT BEACH NUCLEAR PLANT PERIODIC TEST FIRE PROTECTION PROCEDURE UNIT 1 CONTAINMENT FIRE DETECTOR TEST (REFUELING) 1-PT-FP-001 SAFETY RELATED Revision 3 REMARKS: __________________________
_ Page 7 of 11 REFERENCE USE POINT BEACH NUCLEAR PLANT PERJODIC TEST FIRE PROTECTION PROCEDURE UNIT I CONTAINMENT FIRE DETECTOR TEST (REFUELING)
ATTACHMENT A DETECTOR RESPONSE 1-PT-FP-001 SAFETY RELATED Revision 3 Detector Location Acceptance Criteria Initials Notes XS-5175 8' Alarmed XS-5176 8' Alarmed XS-5177 8' Alarmed XS-5178 8' Alarmed XS-5179 8' Alarmed XS-5180 8' Alarmed XS-5181 8' Alarmed XS-5182 8' Alarmed XS-5183 8' Alarmed XS-5184 8' Alarmed XS-5185 21' Alarmed XS-5186 21' Alarmed XS-5187 21' Alarmed XS-5188 21' Alarmed XS-5189 21' Alarmed XS-5190 21' Alarmed Above Blowdown Isolation XS-5191 21' Alarmed Above Blowdown Isolation XS-5192 21' Alarmed Above Fan Cooler lWlCl &2 XS-5193 21' Alarmed Above Fan Cooler 1 Wl Cl & 2 XS-5194 21' Alarmed XS-5195 21' Alarmed XS-5196 21' Alarmed XS-5197 21' Alarmed Above Fan Cooler lWlDl & 2 XS-5198 21' Alarmed XS-5199 21' Alarmed XS-5200 21' Alarmed XS-5201 21' Alarmed XS-5202 21' Alarmed XS-5203 21' Alarmed XS-5204 21' Alarmed From 46' access to Incore Detector XS-5205 46' Alarmed XS-5206 46' Alarmed XS-5207 46' Alarmed Top access from RCP Cubicle XS-5208 46' Alarmed XS-5209 46' Alarmed XS-5210 46' Alarmed XS-5211 46' Alarmed XS-5212 46' Alarmed RCP cubicle top access XS-5213 46' Alarmed XS-5214 46' Alarmed Acceptance Criteria Satisfied for all tested detectors in Attachment A. (Circle one) SAT UNSAT I I Performer Date/Time SRO Date/Time Page 8 of 11 REFERENCE USE POINT BEACH NUCLEAR PLANT PERIODIC TEST FIRE PROTECTION PROCEDURE UNIT 1 CONTAINMENT FIRE DETECTOR TEST (REFUELING) -o** ATTACHMENT B DETECTOR LAYOUT <-~ 0::: ----':J. '. . 0 .* :z: ' .. *.} . . ... Page 9 of 11 l-PT-FP-001 SAFETY RELATED Revision 3 .5 I *,. *-' 0
* I "' ul* ..: 5 " REFERENCE USE 
""'1:1 p) (!Q (1) ....... 0 0 H) ....... ....... NSB 0 -~--. NORTH. ' /* \'. r XS 5197 ., 5198(1.Cl UNIT 1 CONT. I_EL 26'-0"I ~-> C -COND_LIIT
!ABOVE *cEJLINGI
/ii\ COND.UIT !BELOW .CEILING!
I& DETUTOR/DEVICE FLEXIBLE EDNDUIT -@ >-rj ''l:Po ~->--< tr:I 0 Si z &sect;3 "l:10.....:J t-< 0 :;d v b:l zO o n tr:I oZ >-3 > ;: 0 tr:I 0 >-3 (fl I o.....:Ja z 0 ""'1:1 t-< .....:J :;d tr:I fTj 0 > ...... 0 :;d tr:I ""'1:1 v t-< u >-3 tn n .....:J 0 :;d (fl >-3 POINT BEACH NUCLEAR PLANT PERIODIC TEST FIRE PROTECTION PROCEDURE UNIT I *CONTAIN1\1ENT FIRE DETECTOR TEST (REFUELING)
_,. 8 ,. ~. p t:! / l2 J*" <., _, .. 1, ./ .. .. ,.. , l* .. Page 11 of 11 1-PT-FP-001 SAFETY RELATED Revision 3 . . ( .I i "' r z -' -' B ~j w &sect; " l; 0 '? > 8 !;a! w C, -*!:: w .j ::, ::, X -~ w f-' l!l i ' G// I I @(_ I I } 0 I ::: C ,.: z ,o u !:: ) z '* ::, ,. J ( ./' r REFERENCE USE POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER A. PURPOSE ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 1 of 96 1. This procedure provides directions to respond to a loss of all AC safeguards power. 2. This procedure is applicable for initiating events occurring in MODES 1, 2, 3, and 4. B. SYMPTOMS OR ENTRY CONDITIONS
: 1. The symptom of a loss of all AC power is the indication that both AC safeguards trains are deenergized.
: 2. This procedure is entered from:
* EOP-0 UNIT 1, REACTOR TRIP OR SAFETY INJECTION, Step 3, on indication that both AC safeguards trains are deenergized.
* This procedure may be entered directly, on indication that both AC safeguards trains are deenergized.
C. REFERENCES
: 1. EC 283586, Transition to 10 CFR 50.48(c) -NFPA 805 From App R 2. EPM Report R2168-1003C-001 POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER lsTEPI I ACTION/EXPECTED RESPONSE 0 Verify Reactor Trip: 0
* Reactor trip and bypass breakers -OPEN
* Neutron flux-LOWERING Verify Turbine Trip: a. Turbine stop valves-BOTH SHUT ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 2 of 96 11 RESPONSE NOT OBTAINED I ...___ _____ ____, Manually trip reactor:
* Train A
* Train B a. Manually trip turbine. IF Turbine will not trip,~ SHUT MSIVs. NOTES
* Foldout page shall be monitored throughout this procedure.
* CSF status trees should be monitored for information only. CSPs should not be implemented.
3 Secure RCPs a. Ensure both RCPs -STOPPED
* lP-lA
* lP-lB b. Place steam dump mode control -MANUAL POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 3 of 96 lsTEPI l...._~~A~c_T_I_o_N_I_E_x_P_E_c_T_E_D~RE~s_P_o_N_s_E~~--'''
...... ~~~~-RE~_s_P_o_N_s_E~N_o_T~o_B_T_A_I_NE~D~~~~~_.I 4 Check If RCS Is Isolated:
: a. PZR PORVs -BOTH SHUT
* lRC-430
* 1RC-431C b. Letdown orifice outlet valves -SHUT
* 1CV-200A
* 1CV-200B
* 1CV-200C c. Letdown containment isolation valves -SHUT
* 1CV-371A
* lCV-371 d. RCP seal return isolation valve -SHUT
* 1CV-313A e. RCS sample valves -SHUT
* 1SC-966A, PZR steam space sample containment isolation valve
* 1SC-966B, PZR liquid space sample containment isolation valve
* 1SC-966C, RCS hot leg sample containment isolation valve f. Head vent system -ENERGIZED
: g. Reactor vessel head vent solenoids
-SHUT
* 1RC-570A
* 1RC-570B h. PZR vent valves -SHUT
* 1RC-580A
* 1RC-580B a. IF PZR pressure is less than 2335 psig, THEN manually shut PORVs. b. Manually shut valves. c. Manually shut valves. d. Manually shut valve. e. Manually shut valves. f. Go to Step 5. g. Manually shut valves. h. Manually shut valves.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 4 of 96 jsTEPI l.__~~-A_c_T_I_o_N_f_E_x_P_E_c_T_E_D,~RE~s_P_o_N_s_E~~--'11.__~~~~~RE~s-P_o_N_s_E~N-o_T~o-B_T_A_I_NE~D~~~~~....,I 5 Verify AFW Flow -GREATER THAN OR EQUAL TO 230 gpm. Perform the Following:
: a. Ensure TDAFW Pump steam supply MOV's OPEN:
* lMS-2020
* lMS-2019 b. Ensure TDAFW Pump discharge MOV's OPEN:
* lAF-4000
* lAF-4001 c. IF AFW NOT established THEN perform ATTACHMENT K, ESTABLISHING HEAT SINK while continuing with this procedure.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER I STEP' I 6 ACTION/EXPECTED RESPONSE TRY TO RESTORE POWER TO ANY SAFEGUARDS BUS a. Emergency Diesel Generators
-All Running:
* G-01, train A
* G-02, train A (alternate)
* G-03, train B
* G-04, train B (alternate)
: b. AC safeguards Buses -AT LEAST ONE TRAIN ENERGIZED o lA-05 AND lB-03 OR o lA-06 AND lB-04 c. Go to Step 9 II ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 5 of 96 RESPONSE NOT OBTAINED a. Try to start non-running EDGs b. 1) Ensure diesel mode selector switch in AUTO. 2) Place control switch to START . 3) Ensure generator field flash occurs. 4) Ensure green READY TO LOAD light is energized.
: 5) IF NO diesel is running, THEN go to Step 10. IF NO 4160v SAFEGUARDS BUS is energized, THEN go to STEP 7. 1) IF Bus lA-05 is energized, THEN energize lB-03 by closing:
* 1A52-58
* 1B-16B Go to Ste_e 9 2) IF Bus lA-06 is energized, THEN energize lB-04 by closing:
* 1A52-84
* 1B52-17B Go to Ste12 9 I POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL Ac POWER ACTION/EXPECTED RESPONSE II CAUTION ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 6 of 96 RESPONSE NOT OBTAINED In fire scenarios, G-02 is susceptible to automatic tripping on overload when it is supplying both lA-05 and 2A-05 due to spurious equipment operations.
7 Restore Power to "A" Safeguards Bus: a. Check G-01 -RUNNING: 1) IF G-01 is running~ 1A52-60, G-01 to Bus lA-05 Breaker, is NOT closed, THEN perform the following:
a) Ensure 1A52-57, lA-03 to lA-05 Bus Tie Breaker, is OPEN. b) Try to auto close breaker by placing control switch to trip position then release. c) IF breaker will NOT auto close, THEN perform the following:
: 1. Place mode selector switch in EXERCISE 2. Turn synch switch ON 3. At C-02, manually CLOSE breaker control switch 4. Turn synch switch OFF d) ,!,! lA-05 is energized from its normal EOG G-01, THEN energize lB-03 by closing:
* 1A52-58
* 1B52-16B e) Go to Step 9. a. Check G-02 -RUNNING 1) IF G-02 is running AND lA-05 is still NOT energized, THEN perform the following:
a) Ensure 1A52-57, lA-03 to lA-05 Bus Tie Breaker, is OPEN b) Ensure 1A52-60, G-01 to Bus lA-05 breaker, is OPEN and in PULLOUT c) Unlock and place 1A52-66, G-02 to Bus lA-05 breaker, control switch in auto d) IF 1A52-66, G-02 to Bus lA-05 breaker, is NOT closed, THEN perform the following:
: 1) Try to auto close breaker by placing control switch to trip position then release 2) IF breaker will NOT auto close, THEN perform the following:
: a. Place mode selector switch in EXERCISE b. Turn synch switch ON c. At C-02, manually CLOSE breaker control switch d. Turn synch switch OFF e) ,!,! lA-05 is energized from its alternate EOG G-02, THEN energize lB-03 by CLOSING:
* 1A52-58
* 1B52-16B f) Go to Step 9. I POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ACTION/EXPECTED RESPONSE II CAUTION ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 7 of 96 RESPONSE NOT OBTAINED In fire scenarios, G-04 is susceptible to automatic tripping on overload when it is supplying both lA-06 and 2A-06 due to spurious equipment operations.
8 Restore Power to "B" Safeguards Bus: a. Check G-03 -RUNNING: 1) IF G-03 is running~ 1A52-80, G-03 to Bus lA-06 Breaker, is NOT closed, THEN perform the following:
a) Ensure 1A52-77, lA-04 to lA-06 Bus Tie Breaker, is OPEN. b) Try to auto close breaker by placing control switch to trip position then release. c) IF breaker will NOT auto close, THEN perform the following:
: 1. Place mode selector switch in EXERCISE 2. Turn synch switch ON 3. At C-02, manually CLOSE breaker control switch 4. Turn synch switch OFF d) IF lA-06 is energized from its normal EOG G-03, THEN energize lB-04 by closing:
* 1A52-84
* 1B52-17B e) Go to Step 9. a. Check G-04 -RUNNING 1) IF G-04 is running AND lA-06 is still NOT energized, THEN perform the following:
a) Ensure 1A52-77, lA-04 to lA-06 Bus Tie Breaker, is OPEN b) Ensure 1A52-80, G-03 to Bus iA-06 breaker, is OPEN and in PULLOUT c) Unlock and place G-04 to Bus lA-06 breaker 1A52-86 control switch in auto d) IF 1A52-86, G-04 to Bus lA-06 breaker, is NOT closed, THEN perform the following:
e) f) 1) Try to auto close breaker by placing control switch to trip position then release 2) IF breaker will NOT IF -auto close, THEN perform the following:
: a. Place mode selector switch in EXERCISE b. Turn synch switch ON c. At C-02, manually CLOSE breaker control switch d. Turn synch switch OFF lA-06 is energized from its alternate EDG G-04, THEN energize lB-04 by CLOSING:
* 1A52-84
* 1B52-17B Go to Step 9. I POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 8 of 96 E:J 1 ....... ___ A_c_T_I_o_N_f_E_x_P_E_c_T_E_n_RE_s_P_o_N_s_E
___ ..... 11 ....... _____ RE_s_P_o_N_s_E
__ N_o_T_o_B_T_A_I_NE_n
_____ __.l 9 Verify One Train Of Safeguards Bus Energized:
: a. AC safeguards Buses -AT LEAST ONE TRAIN ENERGIZED o lA-05 AND lB-03 OR o lA-06 AND lB-04 b. Monitor running EOG status:
* Check frequency on running diesels -BETWEEN 59.7 Hz AND 60.3 Hz
* Check voltage on running diesels -BETWEEN 4050 Vac AND 4300 Vac c. Service water header pressure GREATER THAN OR EQUAL TO 50 psig a. Try to energize AC safeguards Bus from Control Room using any available power source. Go to Step 10. b. Perform the following:
* Adjust governor to establish 60 Hz
* Adjust voltage regulator to establish 4160 VAC IF EOG cooling, voltage, or frequency can NOT be maintained, THEN shutdown affected EDG(s) 1) Trip affected EDG(s) and place output breaker in pull out: o G-01, 1A52-60 o G-02, 1A52-66 o G-03, 1A52-80 o G-04, 1A52-86 2) Place mode selector switch for affected EDG(s) to LOCAL 3) IF any EOG can NOT be shut down from the Control Room, THEN locally push affected EDG(s) engine stop push button c. Manually start pumps and align valves as necessary to establish service water header pressure greater than or equal to 50 psig. (Step 9, continued on next page)
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 69 LOSS OF ALL AC POWER Page 9 of 96 lsTEPI I ACTION/EXPECTED RESPONSE II RESPONSE NOT OBTAINED I (Step 9. continued from previous page) d. Trip and close contactor(s) for tripped battery chargers aligned to supply DC Buses ,, 0 D-07 0 D-08 0 D-09 0 D-107 0 D-108 0 D-109 e. Return to procedure and step in effect and implement CSPs as necessary CAUTION Local manual closure of breakers bypasses all lockouts and interlocks and may result in equipment damage. 10 Check lA-03: Go to Step 28.
* No Lockouts 11 Check lA-05: Go to Ste:e 28.
* No Lockouts 12 Check lX-04 Go to Ste:e 18.
* No Lockouts
* Power Available 0 G-05 GTG 0 Offsite POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 10 of 96 lsTEPI l~~~A~c_T_I_o_N_I_E_x_P_E_c_T_E_D~RE~s_P_o_N_s_E~~__.11
..... ~~~~-RE~-s-P_o_N_s_E~N-o_T~o-B_T_A_I_NE~D~~~~~
.... 1 13 Restore Power To 13.8 kV: 14 a. Check bus H-02 -ENERGIZED Energize Bus lA-03 From lX-04 a. Reset and CLOSE Bus H-02 feed to lX-04
* H52-22 b. Ensure lA-03 to lA-01 Bus Tie Breaker -OPEN
* 1A52-37 c. Reset and CLOSE Bus lA-03 Normal Feed
* 1A52-36 a. Perform the following:
IF Bus H-01 is energized OR Bus H-03 energized,~:
: 1) Ensure Bus H-02 Normal Feed -OPEN
* H52-20 2) *Ensure H-03 to H-01 Bus Tie Breaker -CLOSED
* H52-31 3) CLOSE H-02 to H-01 Bus Tie Breaker.
* H52-21 IF Bus H-01 is~ energized, THEN start G-05 Gas Turbine per ATTACHMENT E, POWER RESTORATION USING GAS TURBINE, Continue with Step 18. Go to Step 18.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 11 of 96 jsTEPj l...._~~A~c_T_I_o_N_I_E_x_P_E_c_T_E_D~RE~s-P_o_N_s_E~~__,'L-'~~~~-RE~_s_P_o_N_s_E~N_o_T~o_B_T_A_I_NE~D~~~~~-1' 15 16 Energize Bus lA-05 From lA-03 a. Ensure G-01 to Bus lA-05 breaker OPEN
* 1A52-60 b. Ensure G-02 to Bus lA-05 breaker OPEN
* 1A52-66 c. Turn on synchronizing switch for lA-03 to lA-05 Bus Tie Breaker
* 1A52-57 d. Trip and close lA-03 to lA-05 Bus Tie Breaker
* 1A52-57 e. Turn off synchronizing switch for lA-03 to lA-05 Bus Tie Breaker
* 1A52-57 Check 480 V Safeguard Bus lB-03 -ENERGIZED 17 Go to Step 46. 18 Check lX-02
* No Lockouts
* Power Available Go to Step 28. Perform the following:
: a. Close Bus lA-05 feed to lX-13
* 1A52-58 b. Close Bus lB-03 Normal Feed
* 1B52-16B IF lB-03 NOT energized, THEN go to Step 28. Go to Step 23.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 12 of 96 lsTEPI l.._~~A~c_T_I_o_N_f_E_x_P_E_c_T_E_D~RE~s-P_o_N_s_E~~__.ll._~~~~-RE~-s-P_o_N_s_E~N-o_T~o-B_T_A_I_NE~D~~~~~
.... 1 19 20 21 Energize Bus lA-03 From lA-01 a. Ensure Bus lA-03 Normal Feed -OPEN
* 1A52-36 b. Ensure lA-03 to 2A-03 Bus Tie Breaker -OPEN
* 1A52-40 c. Turn synch switch for lA-03 to lA-01 Bus Tie Breaker -ON
* 1A52-37 d. Close lA-03 to lA-01 Bus Tie Breaker
* 1A52-37 e. Turn synch switch for lA-03 to lA-01 Bus Tie Breaker -OFF
* 1A52-37 Energize Bus lA-05 From lA-03 a. Ensure G-01 to Bus lA-05 breaker OPEN
* 1A52-60 b. Ensure G-02 to Bus lA-05 breaker OPEN
* 1A52-66 c. Turn on synchronizing switch for lA-03 to lA-05 Bus Tie Breaker
* 1A52-57 d. Trip and close lA-03 to lA-05 Bus Tie Breaker
* 1A52-57 e. Turn off synchronizing switch for lA-03 to lA-05 Bus Tie Breaker
* 1A52-57 Check 480 V Safeguard Bus lB-03 -ENERGIZED Go to Step 23. Go to Step 26. Perform the following:
: a. Close Bus lA-05 feed to lX-13
* 1A52-58 b. Close Bus lB-03 Normal Feed
* 1B52-16B IF lB-03 NOT energized, THEN go to Step 28.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 13 of 96 lsTEPI l~~~A~c_T_I_o_N_I_E_x_P_E_c_T_E_D~RE~s_P_o_N_s_E~~~ll
..... ~~~~-RE~-s_P_o_N_s_E~N_o_T~o_B_T_A_I_NE~D~~~~~
.... 1 22 Go To Step 46. 23 24 25 Check 2X-04
* No Lockouts
* Power Available Energize Bus lA-03 From 2A-03 a. Ensure Bus lA-03 Normal Feed -OPEN
* 1A52-36 b. Ensure lA-03 to lA-01 Bus Tie Breaker -OPEN
* 1A52-37 c. Turn synch switch for lA-03 to 2A-03 Bus Tie Breaker -ON
* 1A52-40 d. Close lA-03 to 2A-03 Bus Tie Breaker
* 1A52-40 e. Turn synch switch for lA-03 to 2A-03 Bus Tie Breaker -OFF
* 1A52-40 Energize Bus lA-05 From lA-03 a. Ensure G-01 to Bus lA-05 breaker OPEN
* 1A52-60 b. Ensure G-02 to Bus lA-05 breaker OPEN
* 1A52-66 c. Turn on synchronizing switch for lA-03 to lA-05 Bus Tie Breaker
* 1A52-57 d. Trip and close lA-03 to lA-05 Bus Tie Breaker
* 1A52-57 e. Turn off synchronizing switch for lA-03 to lA-05 Bus Tie Breaker
* 1A52-57 Go to Step 28. Go to Step 28. Go to Step 28.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 14 of 96 I STEP 11 ACTION/EXPECTED RESPONSE 11 RESPONSE NOT OBTAINED I 26 27 28 29 30 31 Check 480 V Safeguard Bus lB-03 -ENERGIZED Go To Ste.e_ 46. Check lA-04:
* No Lockouts Check lA-06:
* No Lockouts Check lX-04
* No Lockouts
* Power Available 0 G-05 GTG 0 Offsite Restore Power To 13.8 kV: a. Check bus H-02 -ENERGIZED Perform the following:
: a. Close Bus lA-05 feed to lX-13
* 1A52-58 b. Close Bus lB-03 Normal Feed
* 1B52-16B IF lB-03 NOT energized, THEN go to Step 28. Go to Step 46. Go to Step 46. Go to Step 36 a. Perform the following:
IF Bus H-01 is energized OR Bus H-03 energized,~:
: 1) Ensure Bus H-02 Normal Feed -OPEN
* H52-20 2) Ensure H-03 to H-01 Bus Tie Breaker -CLOSED
* H52-31 3) CLOSE H-02 to H-01 Bus Tie Breaker.
* H52-21 IF Bus H-01 is NOT energized, THEN start G-05 Gas Turbine per ATTACHMENT E, POWER RESTORATION USING GAS TURBINE, Continue with Step 36.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 15 of 96 lsTEPI l._~~-A_c_T_I_o_N~/_E_x_P_E_c_T_E_D~RE~s_P_o_N~S-E~~__.IIL-~~~~~RE~s-P_o_N_s_E~-N-o_T~o-B_T_A_I_NE~-D~~~~~...11 32 33 34 Energize Bus lA-04 From lX-04 a. Reset and CLOSE Bus H-03 feed to lX-04
* H52-22 b. Ensure lA-04 to lA-02 Bus Tie Breaker -OPEN
* 1A52-55 c. Reset and CLOSE Bus lA-04 Normal Feed
* 1A52-56 Energize bus lA-06 From lA-04 a. Ensure G-03 to Bus lA-06 breaker OPEN
* 1A52-BO b. Ensure G-04 to Bus lA-06 breaker OPEN
* 1A52-86 c. Trip and CLOSE Bus lA-04 Normal Feed to lA-06
* 1A52-54 d. Turn on synchronizing switch for lA-04 to lA-06 Bus Tie Breaker
* 1A52-77 e. Trip and close lA-04 to lA-06 Bus Tie Breaker
* 1A52-77 f. Turn off synchronizing switch for lA-04 to lA-06 Bus Tie Breaker
* 1A52-77 Check 480 V Safeguard Bus lB-04 -ENERGIZED 35 Go to Step 46. Go to Step 36. Go to Step 46. Perform the following:
: a. Close Bus lA-06 feed to lX-14
* 1A52-84 b. Close Bus lB-04 Normal Feed
* 1B52-17B IF lB-04 NOT energized, THEN go to Step 46.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 16 of 96 I STEP 11 ACTION/EXPECTED RESPONSE 11 RESPONSE NOT OBTAINED I 36 37 38 Check lX-02
* No Lockouts
* Power Available Energize Bus lA-04 From lA-02 a. Ensure Bus lA-04 Normal Feed -OPEN
* 1A52-56 b. Ensure lA-04 to 2A-04 Bus Tie Breaker -OPEN
* 1A52-52 c. Turn synch switch for lA-04 to lA-02 Bus Tie Breaker -ON
* 1A52-55 d. Close lA-04 to lA-02 Bus Tie Breaker
* 1A52-55 e. Turn synch switch for lA-04 to lA-02 Bus Tie Breaker -OFF
* 1A52-55 Energize bus lA-06 From lA-04 a. Ensure G-03 to Bus lA-06 breaker OPEN
* 1A52-80 b. Ensure G-04 to Bus lA-06 breaker OPEN
* 1A52-86 c. Trip and CLOSE Bus lA-04 Normal Feed to lA-06
* 1A52-54 d. Turn on synchronizing switch for lA-04 to lA-06 Bus Tie Breaker
* 1A52-77 e. Trip and close lA-04 to lA-06 Bus Tie Breaker
* 1A52-77 f. Turn off synchronizing switch for lA-04 to lA-06 Bus Tie Breaker
* 1A52-77 Go to Step 41. Go to Step 41. Go to Step 46.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 17 of 96 ~-I ___ A_c_T_I_o_N_I_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
__ ___.I I ______ RE_. _s_P_oN_s_E_N_o_T_o_B_T_A_I_NE_D
_____ .... 1 39 Check 480 V Safeguard Bus lB-04 -ENERGIZED 40 Go to Step 46. 41 42 Check 2X-04
* No Lockouts
* Power Available Energize Bus lA-04 From 2A-04 a. Ensure Bus lA-04 Normal Feed -OPEN
* 1A52-56 b. Ensure lA-04 to lA-02 Bus Tie Breaker -OPEN
* 1A52-55 c. Turn synch switch for lA-04 to 2A-04 Bus Tie Breaker -ON
* 1A52-52 d. Close lA-04 to 2A-04 Bus Tie Breaker
* 1A52-52 e. Turn synch switch for lA-04 to 2A-04 Bus Tie Breaker -OFF
* 1A52-52 Perform the following:
: a. Close Bus lA-06 feed to lX-14
* 1A52-84 b. Close Bus lB-04 Normal Feed
* 1B52-17B IF lB-04 NOT energized, THEN go to Step 46. Go to Step 46. Go to Step 46.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 18 of 96 I STEP 11 ACTION/EXPECTED RESPONSE 11 RESPONSE NOT OBTAINED I ....___ ____ __,....._
_____ ___. 43 44 Energize bus lA-06 From lA-04 a. Ensure G-03 to Bus lA-06 breaker OPEN
* 1A52-80 b. Ensure G-04 to Bus lA-06 breaker OPEN
* 1A52-86 c. Trip and CLOSE Bus lA-04 Normal Feed to lA-06
* 1A52-54 d. Turn on synchronizing switch for lA-04 to lA-06 Bus Tie Breaker
* 1A52-77 e. Trip and close lA-04 to lA-06 Bus Tie Breaker
* 1A52-77 f. Turn off synchronizing switch for lA-04 to lA-06 Bus Tie Breaker
* 1A52-77 Check 480 V Safeguard Bus lB-04 -ENERGIZED 45 Go to Step 46. Go to Step 46. Perform the following:
: a. Close Bus lA-06 feed to lX-14
* 1A52-84 b. Close Bus lB-04 Normal Feed
* 1B52-17B IF lB-04 NOT energized, THEN go to Step 46.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 19 of 96 lsTEPI I ACTION/EXPECTED RESPONSE II RESPONSE NOT OBTAINED I ,_ ____ __,....._
_____ ____. 46 Verify One Train Of Safeguards Bus Energized:
: a. AC safeguards Buses -AT LEAST ONE TRAIN ENERGIZED o lA-05 AND lB-03 OR o lA-06 AND lB-04 b. Service water header pressure GREATER THAN OR EQUAL TO 50 psig c. Trip and close contactor(s) for tripped battery chargers aligned to supply DC Buses o D-07 o D-08 o D-09 o D-107 o D-108 o D-109 d. Return to Procedure and Step in effect and implement CSPs as necessary
: a. Perform the following:
: 1) Locally monitor AFW Pump room temperatures.
IF AFW pump room temperature rises to 120&deg;F or greater, THEN refer to AOP-30, Temporary Ventilation for Vital Areas 2) OBSERVE CAUTIONS PRIOR TO STEP 47 and go to Step 47. b. Manually start pumps and align valves as necessary to establish service water header pressure greater than or equal to 50 psig. 
 
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ~I ACTION/EXPECTED RESPONSE (Step 47. continued from previous page)
* NON-running condensate pumps: 0 1P-25A 0 1P-25B
* NON-running circulating water pumps: 0 1P-30A 0 1P-30B
* Containment accident fans:
* lW-lAl
* lW-lBl
* lW-lCl
* lW-101
* Containment spray pumps:
* 1P-14A
* 1P-14B
* SI pumps:
* 1P-15A
* 1P-15B II ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 21 of 96 RESPONSE NOT OBTAINED I POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ACTION/EXPECTED RESPONSE II NOTE ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 22 of 96 RESPONSE NOT OBTAINED Refer to ATTACHMENT J for a list of equipment that should be available during a loss of all AC power. 48 Try To Restore Power To Any 480 Vac Safeguards Bus While Continuing With This Procedure 0 Locally start GOl per ATTACHMENT A 0 Locally start G02 per ATTACHMENT B 0 Locally start G03 per ATTACHMENT C 0 Locally start G04 per ATTACHMENT D 0 Locally start G05 per ATTACHMENT E 0 Backfeed 480 Vac buses per ATTACHMENT F 49 Isolate RCP Seals: a. Shut at least one RCP seal return containment isolation valve: o 1CV-313A o lCV-313 b. Locally shut the following valves outside containment:
* RCP seal injection throttle valves:
* 1CV-300A
* 1CV-300B
* RCP component cooling return isolation valves:
* 1CC-759A
* 1CC-759B Align equipment to alternate power source per ATTACHMENT G, ALIGNING EQUIPMENT TO ALTERNATE POWER SOURCE, while continuing with this procedure.
: a. Locally shut RCP seal water return isolation valve outside containment:
* lCV-313 I POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ACTION/EXPECTED RESPONSE II NOTE ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 23 of 96 RESPONSE NOT OBTAINED ELAP is short for Extended Loss of AC Power. It is longer than SBO coping time. I **********************************************************
50 Check If AC Power Can Be Restored -WITHIN ONE HOUR AC power can NOT be restored in one hour, THEN perform the following while continuing on with this procedure:
* FSG-5, INITIAL ASSESSMENT AND FLEX EQUIPMENT STAGING
* FSG-4, ELAP DC BUS LOAD SHED/MANAGEMENT
**********************************************************
51 Check CST Isolated From Condenser Hotwell: a. Ensure condenser hotwell low flow make-up valve -SHUT
* lCS-2125 b. Ensure condenser manual fill valve -SHUT
* lCS-86 a. IF valve CANNOT be manually shut, THEN locally shut upstream isolation valve:
* lCS-92 POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 24 of 96 I STEP 11 ACTION/EXPECTED RESPONSE 11 RESPONSE NOT OBTAINED I 52 Check S/G Status: a. MSIVs -BOTH SHUT
* lMS-2018
* lMS-2017 b. MSIV bypass valves -BOTH SHUT
* lMS-234
* lMS-236 c. Feedwater isolation valves -BOTH SHUT
* lCS-3124
* lCS-3125 d. Blowdown isolation valves -SHUT
* lMS-5958
* lMS-5959 53 Check S/G's Available For Cooldown:
: a. Both S/G ADV's are: o Accessible o Capable of operating
: a. Manually shut MSIVs. IF any MSIVs CANNOT be shut, THEN locally shut MSIV(s) per ATTACHMENT H. b. Locally shut valves. c. Perform the following:
: 1) Shut feedwater regulating valves:
* 1FIC-466A
* 1FIC-476A
: 2) Shut feedwater regulating bypass valves:
* lCS-480
* lCS-481 d. Shut blowdown header isolation valves:
* lMS-2042
* lMS-2045 a. IF an asymmetrical cooldown is required, THEN go to ATTACHMENT M, Asymmetrical Cooldown.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ACTION/EXPECTED RESPONSE II CAUTIONS ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 25 of 96 RESPONSE NOT OBTAINED
* A faulted or ruptured S/G that is isolated should remain isolated.
* Alternate AFW pump suction supply will be necessary if CST level lowers to less than 4 ft. NOTE If an ELAP is in progress and CST level lowers to 15.75 ft., FSG-6, ALTERNATE CST MAKEUP should be used if CSTs are available I **********************************************************
54 Check Intact S/G Levels: a. S/G levels -GREATER THAN [51%] 32% a. Maintain maximum AFW flow until level is greater than [51%] 32% in at least one S/G. IF an ELAP is NOT in progress, THEN perform ATTACHMENT K, ESTABLISHING HEAT SINK. IF an ELAP is in progress AND CST is NOT available, THEN perform FSG-2, ALTERNATE AFW SUCTION SOURCE an ELAP is in progress AND an alternate low pressure feedwater source is required due to lP-29 unavailability, THEN perform the following:
: 1. IF 2P-29 AND associated suction source is available, THEN perform FSG-15 UNIT 1, CROSS TIE AFW. 2. IF 2P-29 AND associated suction source NOT available, THEN establish low pressure feedwater per ATTACHMENT L, ESTABLISHING LOW PRESSURE FEEDWATER FLOW. ********************************************************** (Step 54, continued on next page)
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 26 of 96 lsTEPj l._~~-A_c_T_I_o_N_f_E_x~P-E_c_T_E_D~RE~s_P_o_N~s_E~~--'''--~~~~~RE~s-P_o_N_s_E~-N-o_T~o-B_T_A_I_NE~D~~~~~~I (Step 54. continued from previous page) **********************************************************
: b. Control AFW flow to maintain S/G levels between [51%] 32% and 63% b. IF level in any S/G continues to rise in an uncontrolled manner, THEN isolate ruptured S/G (s): 1) Reset Loss Of Feedwater Turbine Trip 2) SHUT MDAFW pump discharge valves: 0 1AF-4074A for S/G A 0 1AF-4074B for SIG B 3) Shut turbine driven AFW pump discharge valves: 0 lAF-:-4001 for S/G A 0 lAF-4000 for S/G B 4) Shut steam supply to turbine-driven AFW pump: 0 lMS-2020 for S/G A 0 lMS-2019 for S/G B 5) Adjust ruptured S/G(s) atmospheric steam dump controller to 1050 psig o lHC-468 for S/G A o lHC-478 for S/G B 6) WHEN ruptured S/G(s) pressure is less than 1050 psig, THEN ensure atmospheric steam dump is shut: o lMS-2016 for S/G A o lMS-2015 for S/G B .!,! atmospheric steam dump CANNOT be shut, THEN locally isolate atmospheric steam dump: o lMS-227 for S/G A o lMS-244 for S/G B **********************************************************
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 27 of 96 lsTEPI '~~~A~c_T_I_o_N_I_E_x_P_E_c_T_E_D~RE~s_P_o_N_s_E~~~ll..._~~~~-RE~-s_P_o_N_s_E~N-o_T~o_B_T_A_I_NE~D~~~~~
.... 1 55 CAUTION Steam supply to the turbine-driven AFW pump must be maintained from at least one S/G. Check If S/G are NOT faulted:
* NO S/G pressure lowering in an uncontrolled manner AND
* NO S/G completely depressurized Isolate faulted S/G(s): a. Reset Loss Of Feedwater Turbine Trip. b. Shut motor driven AFW pump discharge valves: o 1AF-4074A for S/G A o 1AF-4074B for S/G B c. Shut turbine driven AFW pump discharge valves: o lAF-4001 for S/G A o lAF-4000 for S/G B d. Shut steam supply to turbine-driven AFW pump: o lMS-2020 for S/G A o lMS-2019 for S/G B e. Ensure atmospheric steam dump is shut: o lMS-2016 for S/G A o lMS-2015 for S/G B POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 28 of 96 E:J IL._ ___ A_c_T_I_o_N_f_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
__ __.I IL-_____ RE_s_P_o_N_s_E_N_o_T_o_B_T_A_INE
__ o _____ _.l 56 Check If S/G Tubes Are NOT ruptured:
: a. Secondary system radiation monitor levels -NORMAL
* Condenser air ejector:
* lRE-215 (3-5)
* RE-225 (7-1)
* SIG blowdown:
* lRE-219 (5-3)
* lRE-222 (1-7)
* Main steam line:
* lRE-231 (3-9) for SIG A
* lRE-232 (5-2) for SIG B b. Request Chemistry to periodically sample both SIGs for activity c. Request local surveys of main steam lines b. Secondary activity samples and surveys -NORMAL (WHEN AVAILABLE)
Perform the following:
: 1. Try to identify ruptured SIG(s) while continuing with this procedure.
: 2. WHEN ruptured SIG(s) identified, THEN isolate ruptured SIG(s): a) Reset Loss Of Feedwater Turbine Trip. b) Shut motor driven AFW pump discharge valves: o 1AF-4074A for SIG A o 1AF-4074B for SIG B c) Shut turbine driven AFW discharge valves: 0 lAF-4001 for SIG A 0 lAF-4000 for SIG B d) Shut steam supply to turbine-driven AFW pump: o lMS-2020 for SIG A o lMS-2019 for SIG B e) Adjust ruptured SIG(s) atmospheric steam dump controller to 1050 psig: o lHC-468 for SIG A o lHC-478 for SIG B f) ruptured SIG(s) pressure is less than 1050 psig, THEN ensure pump atmospheric steam dump is shut: 0 lMS-2016 for SIG A 0 lMS-2015 for SIG B POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 29 of 96 jsTEPI...._
'~~-A_c_T_I_o_N_I_E_x_P_E_c_T_E_D~RE~s_P_o_N_s_E~~___.ll.._~~~~-RE~-s-P_oN~S-E~N-o_T~o-B_T_A_I_NE~D~~~~~-''
57 Check DC Bus Loads: a. Check ELAP -IN PROGRESS b. Check vital instrumentation
-AVAILABLE
* Any DC Bus Voltage GREATER THAN 107.5 Vdc
* Vital instruments
-REQUIRED INSTRUMENTS AVAILABLE o Red Instrument Bus Powered o White Instrument Bus Powered 58 Check CST Level a. CST Level -GREATER THAN 15.75 ft. b. CST Level -GREATER THAN 4 ft. a. Go to Step 58. b. Perform FSG-7 Unit 1, LOSS OF VITAL INSTRUMENTATION OR CONTROL POWER while continuing on with this procedure.*
: a. IF an ELAP is in progress AND CST is Available, THEN perform FSG-6, ALTERNATE CST MAKEUP while continuing with this procedure
: b. IF an ELAP NOT in progress, THEN switch to alternate AFW suction supply per AOP-23, UNIT 1, ESTABLISHING ALTERNATE AFW SUCTION SUPPLY while continuing with this procedure.
IF an ELAP in progress, THEN perform FSG-2, ALTERNATE AFW SUCTION SOURCE while continuing with this procedure POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ~, ACTION/EXPECTED RESPONSE 11 NOTE ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 30 of 96 RESPONSE NOT OBTAINED RCS temperature should be stabilized prior to taking action based on Low PZR level. This may require controlling steam generator feed rate and allowing temperature to return to 547&deg;F. I **********************************************************
59 Monitor RCS Integrity
: a. Monitor RCS inventory-'
MAINTAINED AS EXPECTED
* RCS subcooling based on CET's -Greater than [74&deg;F] 35&deg;F
* PZR Level -Greater than [32%] 13% b. Check plant conditions
-ELAP IN PROGRESS c. Check S/G ADVs -REMOTE CONTOL AVAILABLE
: a. Go to Step 60. b. Perform the following:
: 1) Consult with TSC to determine when plant cooldown should be initiated.
: 2) plant cooldown is desired,~
go to Step 60. Continue with Step 62. c. WHEN RCS cooldown is desired AND personnel are available to manually operate the ADVs, THEN go to Step 60. **********************************************************
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 31 of 96 E=J I...._ __ A_c_T_I_o_N_I_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
__ ___.I I.._ _____ RE __ s_P_o_N_s_E_N_o_T_o_B_T_A_I_NE_o
_____ ....,I CAUTIONS
* S/G pressures should be maintained greater than 280 psig to prevent injection of accumulator nitrogen into the RCS.
* S/G level should be maintained greater than [51%] 32% in at least one intact S/G. If level cannot be maintained, S/G depressurization should be stopped until level is restored in at least one S/G. NOTES
* CST level of 4 ft. is based on the decay heat load for 1 hour following a reactor trip without cool down of the RCS. Depressurization should not be started without either an adequate CST level, an alternate AFW suction supply or fire water aligned to CST.
* The S/G's should be depressurized a rate sufficient to maintain a cooldown rate in the RCS cold legs near 100&deg;F/hr.
This will minimize RCS inventory loss while cooling the RCP seals in a controlled manner.
* PZR level may be lost and reactor vessel upper head voiding may occur due to depressurization of S/Gs. Depressurization should not be stopped to prevent these occurrences.
60 Depressurize Intact S/G's To 320 psig: a. S/G levels -GREATER THAN [51%] 32% IN AT LEAST ONE S/G a. Perform the following:
: 1) Maintain maximum AFW flow until level is greater than [51%] 32% in at least one S/G. 2) IF an ELAP is NOT in progress, THEN perform ATTACHMENT K, ESTABLISHING HEAT SINK. (Step 60, continued on next page)
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__ o _____ .... 1 (Step 60. continued from previous page) b. Manually dump steam using S/G ADVs to maintain cooldown rate in RCS cold legs -LESS THAN 100&deg;F/hr.
o lMS-2016 for S/G A o lMS-2015 for S/G B c. RCS cold leg temperatures
-GREATER THAN 340&deg;F d. S/G pressures
-LESS THAN 320 psig. 3) IF an ELAP is in progress AND an alternate low pressure feedwater source is required due to lP-29 unavailability, THEN perform the following:
a) IF 2P-29 AND associated suction source is available, THEN perform FSG-15 UNIT 1, CROSS TIE AFW b) IF 2P-29 AND associated suction source NOT available, THEN establish low pressure feedwater per ATTACHMENT L, ESTABLISHING LOW PRESSURE FEEDWATER FLOW 4) WHEN level is greater than [51%] 32% in at least one S/G, THEN do Steps 60.b through 60.e Continue with Step 61. b. Locally dump steam using atmospheric steam dump. o lMS-2016 for S/G A o lMS-2015 for S/G B c. Perform the following
: 1) Control atmospheric steam dump to stop S/G depressurization.
: 2) Go to Step 61. d. WHEN S/G pressures lower to less than 320 psig, THEN do Step 60.e. Continue with Step 61. (Step 60, continued on next page)
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..... ~~~~~RE~s-P_o_N_s_E~-N-o_T~o-B_T_A_I_NE~D~~~~~--'I (Step 60. continued from previous page) e. Manually control atmospheric steam dumps to maintain S/G pressures at 320 psig:
* lMS-2016 for S/G A
* lMS-2015 for S/G B e. Locally control atmospheric steam dumps to maintain S/G pressures at 320 psig.
* lMS-2016 for S/G A
* lMS-2015 for S/G B IF S/G pressure decreases to less than 320 psig without operation of atmospheric steam dumps or S/G feed, THEN perform FSG-9 UNIT 1 LOW DECAY HEAT TEMPERATURE CONTROL. **********************************************************
61 Check Reactor Subcritical:
: a. Intermediate range channels-ZERO OR NEGATIVE STARTUP RATE * [1N-40B]
* lN-35
* lN-36 b. Source range channels -ZERO OR NEGATIVE STARTUP RATE * [lN-400]
* lN-31
* lN-32 Control atmospheric steam dump to stop S/G depressurization and allow RCS to heat up:
* lMS-2016 for S/G A
* lMS-2015 for S/G B **********************************************************
62 NOTE Depressurization of S/G's will result in SI actuation.
SI should be reset to permit manual loading of equipment on AC safeguards bus. Check SI -ACTUATED WHEN SI is actuated, THEN do Steps 63, 64 and 65. Continue with Step 66. 63 Reset SI POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 34 of 96 I.__ ___ A_c_T_I_o_N_I_E_x_P_E_c_T_E_D_RE_s_P_o_N_sE
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__ D _____ ....,I 64 Verify Containment Isolation:
65 66 67 a. Containment isolation panels "A" and "B" -ALL LIGHTS LIT Check Containment Pressure Recorder -HAS REMAINED LESS THAN 25 psig
* lPR-968
* lPR-969 Check Core Exit Thermocouple Temperatures
-LESS THAN 1200&deg;F
* lTR-OOOOlA
* lTR-000018 Check Plant Conditions
* ELAP in progress
* S/G depressurization to 320 psig -HAS BEEN COMPLETED
: a. Perform the following:
: 1) Manually actuate Containment Isolation.
: 2) IF any containment isolation valve is open, THEN manually shut valve(s).
Refer to ATTACHMENT I, CONTAINMENT ISOLATION VALVES. IF any valve CANNOT be shut, THEN locally shut valve(s).
Perform the following:
: a. Check containment spray actuated:
* Annunciator COl B 2-6, CONTAINMENT SPRAY, lit IF containment spray has NOT actuated, THEN manually actuate containment spray. b. Reset containment spray signal. IF core exit temperatures are greater than 1200&deg;F AND rising, THEN go to SACRG-1, SEVERE ACCIDENT CONTROL ROOM GUIDANCE INITIAL RESPONSE.
Go to Step 70.
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_____ _.l 68 Depressurize Intact S/G's For Maximum Passive Injection:
: a. Check FSG-8 UNIT 1, ALTERNATE RCS BORATION STATUS
* COMPLETE
* MAXIMUM PASSIVE INJECTION DIRECTED b. Check the following:
* Accumulator-pressure indication available
* FLEX PDG is in service per FSG-5, INITIAL ASSESMENT AND FLEX EQUIPMENT STAGING a. IF required boration has been completed per FSG-8 UNIT 1, THEN go to Step 69. IF NOT, THEN Go to Step 70. b. Perform the following:
: 1) WHEN indication and FLEX Equipment is in place, THEN continue with Step 68.c through 68.j. 2) Continue with Step 70. (Step 68, continued on next page)
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...... 1 (Step 68. continued from previous page) c. Check S/G narrow range levels greater than [51%] 32% d. Manually dump steam using S/G atmospheric dump valves to maintain cooldown rate in RCS cold legs -LESS THAN 100&deg;F/hr e. Monitor reactor -SUBCRITICAL
* Intermediate range channels -ZERO OR NEGATIVE START UP RATE
* Source range channels -ZERO OR NEGATIVE START UP RATE c. Perform the following:
: 1) Maintain maximum AFW flow until narrow range level greater then [51%] 32% in at least one S/G 2) IF an ELAP is in progress AND an alternate low pressure feedwater source is required due to lP-29 unavailability, THEN perform the following:
a) IF 2P-29 AND associated suction source is available,~
perform FSG-15 UNIT 1, CROSS TIE AFW. b) IF 2P-29 AND associated suction source NOT available, THEN establish low pressure feedwater per ATTACHMENT L, ESTABLISHING LOW PRESSURE FEEDWATER FLOW. 3) WHEN narrow range level is greater than [51%] 32% in at least one S/G, THEN continue with Step 68.d through 68.j 4) Continue with Step 70. d. Locally dump steam using S/G atmospheric durrip valves: o lMS-2016 o lMS-2015 e. Control S/G atmospheric dump valves to stop S/G depressurization and allow RCS to heat up Go to Step 68.g. (Step 68, continued on next page)
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_____ ___.l (Step 68. continued from previous page) f. Check either of the following conditions
-MET: o SIG pressures
-AT OR LESS THAN 120 psig o Accumulator pressures
-AT OR LESS THAN 320 psig g. Stabilize S/G pressure h. Check accumulator pressure -AT OR LESS THAN 320 PSIG i. Check accumulator pressure stable j. Isolate accumulators using FSG-10 UNIT 1, UNIT 1 PASSIVE RCS INJECTION ISOLATION
: f. Go to Step 68.d. h. Perform the following:
: 1) IF RCS subcooling is less than [74&deg;F] 35&deg;F, THEN go to Step 68. i. 2) Open RCS head vent valves:
* 1RC-570A
* 1RC-570B 3) WHEN accumulator pressure is less than 320 psig OR RCS subcooling is less than [74&deg;F] 35&deg;F, THEN perform the following:
a) Close RCS head vent valves
* 1RC-570A
* 1RC-570B b) Go to Step 68.i. 4) Continue with Step 70. i. IF accumulator pressure continues to lower, THEN control S/G atmospheric dump valves to increase S/G pressure to 280 psig: o lMS-2016 o lMS-2015 j. Control S/G atmospheric dump valves to lower S/G pressure to 320 psig o lMS-2016 o lMS-2015 POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 38 of 96 I STEPI I ACTION/EXPECTED RESPONSE II RESPONSE NOT OBTAINED I ..___ ____ __.....___
_____ __. 69 Depressurize intact S/G's to 150 psig for long term cooling: a. Check required boration per FSG-8 UNIT 1, ALTERNATE RCS BORATION -COMPLETE b. Perform FSG-10 UNIT 1, PASSIVE RCS INJECTION ISOLATION
: c. Verify accumulators
-ISOLATED
* 1SI-841A
* 1SI-841B d. Check S/G narrow range levels greater than [51%] 32% in at least one S/G a. Perform the following:
: 1) WHEN required boration per FSG-8 UNIT 1, is complete, do Steps 69.b through ~-2) Continue with Step 70. c. Perform FSG-10 UNIT 1, PASSIVE RCS INJECTION ISOLATION
: 1) WHEN accumulators are isolated, THEN continue with Step 69.d through 69.g 2) Continue to Step 70. d. Perform the following:
: 1) Maintain maximum AFW flow until narrow range level greater than [51%] 32% in at least one S/G 2) IF an ELAP is in progress AND an alternate low pressure feedwater source is required due to lP-29 unavailability, THEN perform the following:
a) IF 2P-29 AND associated suction source is available, THEN perform FSG-15 UNIT 1, CROSS TIE AFW. b) IF 2P-29 AND associated suction source NOT available, THEN establish low pressure feedwater per ATTACHMENT L, ESTABLISHING LOW PRESSURE FEEDWATER FLOW. 3) WHEN narrow range level is greater than [51%] 32% in at least one S/G, THEN continue with Step 69.e through 69.g. (Step 69, continued on next page)
L POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 39 of 96 lsTEPI l._~~-A_c_T_I_o_N_I_E_x~P-E_c_T_E_D~RE~s-P_o_N~SE~~~
..... ll._~~~~~RE~s-P_o_N_s_E~-N-o_T~o-B_T_A_I_NE~D~~~~~---1 (Step 69. continued from previous page) e. Manually dump steam using S/G ADVs to maintain cooldown rate in RCS cold legs less than 100&deg;F/hr.
* lMS-2016
* lMS-2015 f. Check S/G pressures
-LESS THAN 150 psig g. Manually control S/G ADVs at 150 psig:
* lMS-2016
* lMS-2015 e. Locally dump steam using S/G ADVs:
* lMS-2016
* lMS-2015 f. Perform the following:
: 4) WHEN S/G pressures lower to less than 150 psig, THEN go to Step 69.e. 5) Continue to Step 70. g. Locally control S/G ADVs to maintain S/G pressure at 150 psig IF S/G pressure can NOT be maintained at 150 psig, THEN perform FSG-9 UNIT 1, LOW DECAY HEAT TEMPERATURE CONTROL to stop the uncontrolled cooldown.
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_____ ____. 70 Check If AC Safeguards Power Is Restored:
: a. AC safeguards buses -AT LEAST ONE TRAIN ENERGIZED o lA-05 and lB-03 OR o lA-06 and lB-04 a. Continue to control RCS conditions and monitor plant status: 1) Check status of local actions:
* AC power restoration (Step 48)
* RCP seal isolation (Step 49)
* Other local actions if applicable:
o Gas Turbine (ATTACHMENT E) o AFW pump room cooling (Step 46 RNO 1)) o Condenser hotwell (Step 51) o MSIV and bypass valve closure (Step 52) o Faulted S/G isolation (Step 55) o Ruptured S/G isolation (Step 56) o CST makeup (Step 58) o Containment isolation (Step 64) 2) Periodically check status of spent fuel cooling:
* Spent fuel pool level greater than 62' 8" El. o LI-40A o LI-40B o Installed Wall Tape
* Spent fuel pool temperature less than 120&deg;F.
* IF level less than 62' 8" OR temperature greater than 120&deg;F, dispatch personnel to initiate actions per AOP-8F, LOSS OF SPENT FUEL POOL COOLING. (Step 70, continued on next page)
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__ o _____ _.l (Step 70. continued from previous page) b. Check if any FSGs implemented.
: c. Perform FSG-13 UNIT 1, TRANSITION FROM FLEX EQUIPMENT.
: d. Observe considerations specified in FSG-13 UNIT 1, TRANSITION FROM FLEX EQUIPMENT before starting equipment in subsequent steps and procedures.
: 3) 4) IF TDAFW Pump is operating, gag OPEN Mini-recirc Valve:
* lAF-4002 IF ELAP is in progress, THEN implement the following strategies as needed. o RCS Inventory Control IF pressurizer level is less than [44%] 24% and time and personnel are available, .Q!3: NR RVLIS is less than 25 ft. AND indication of reflux cooling,~
perform FSG-1 UNIT 1, LONG TERM RCS INVENTORY CONTROL. o Boration IF time since event initiation is greater than 24 hours, THEN perform FSG-8 UNIT 1, ALTERNATE RCS BORATION.
o Containment Cooling IF containment temperature greater than 175&deg;F OR containment pressure greater than 48 psig, THEN perform FSG 12 UNIT 1, ALTERNATE CONTAINMENT COOLING. 5) OBSERVE CAUTION PRIOR TO STEP 54 and return to Step 54. b. Go to Step 71.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 42 of 96 I STEP 11 ACTION/EXPECTED RESPONSE 11 RESPONSE NOT OBTAINED I CAUTION The loads placed on the energized AC safeguards bus should not exceed the capacity of the power source. Refer to AOP-22 UNIT 1, EDG LOAD MANAGEMENT, for KW ratings. 71 Verify Service Water System Operation:
: a. Service water header pressure -GREATER THAN OR EQUAL TO 50 psig 72 Restore Battery Chargers:
: a. Trip and close contactor(s) for tripped battery chargers aligned to supply DC buses: 0 D-07 0 D-08 0 D-09 0 D-107 0 D-108 0 D-109 b. Battery charger(s)
-OPERATING
: a. Manually start pumps and align valves as necessary to establish service water header pressure greater than or equal to 50 psig. b. Restart affected battery charger(s) per the following:
0 0-SOP-DC-001, 125 voe SYSTEM, BUS D-01 & COMPONENTS 0 0-SOP-DC-002, 125 voe SYSTEM, BUS D-02 & COMPONENTS 0 0-SOP-DC-003, 125 voe SYSTEM, BUS D-03 & COMPONENTS 0 0-SOP-DC-004, 125 voe SYSTEM, BUS D-04 & COMPONENTS 0 0-SOP-DC-005, 125 voe SYSTEM, SWING BUSES & COMPONENTS 0 0-SOP-DC-006, 125 voe SYSTEM, NON-VITAL BUSES & COMPONENTS POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 43 of 96 lsTEPI l.__~~-A_c_T_I_o_N_I_E~x_P_E_c_T_E_o~RE~s_P_o_N~s_E~~---'11
..... ~~~~~RE~s_P_o_N_s_E~_N_o_T~o_B_T_A_I_NE~o~~~~~__,I 73 Stabilize S/G Pressures:
: a. Manually control atmospheric steam dumps:
* lMS-2016 for S/G A
* lMS-2015 for S/G B 74 Verify Following Equipment Loaded On AC Safeguards Bus: a. Emergency AC lighting and plant alarms: o lB-42 0 lB-32 b. B-33 and B-43 -AT LEAST ONE ENERGIZED
: c. Communications:
: 1) Gai-tronics
: 2) Radio communications
: a. Locally control atmospheric steam dumps.
* lMS-2016 for S/G A
* lMS-2015 for S/G B a. Energize lB-42 or lB-32. b. Energize B-33 or B-43. c. -Perform one of the following:
a) IF B-33 is energized, THEN position transfer switch to B-33:
* B-50 b) IF B-33 is NOT energized, THEN position transfer switch to B-43:
* B-50 POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 44 of 96 I...._ ___ A_c_T_I_o_N_f_E_x_P_E_c_T_E_n_RE_s_P_o_N_s_E
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______ l NOTE If RCP seal cooling was previously isolated, further cooling of the RCP seals will be established by natural circulation cooldown as directed in subsequent procedures.
75 Select Recovery Procedure:
: a. Check RCS subcooling based on core exit thermocouples
-GREATER THAN [74&deg;F] 35&deg;F b. Check PZR level -GREATER THAN [32%] 13% c. Check SI equipment
-HAS NOT ACTUATED UPON AC POWER RESTORATION
: d. Go to ECA-0. 1 UNIT 1, LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED a. b. C. -END-Go to ECA-0. ALL AC POWER REQUIRED.
Go to ECA-0. ALL AC POWER REQUIRED.
Go to ECA-0. ALL AC POWER REQUIRED.
2 UNIT 1, LOSS OF RECOVERY WITH SI 2 UNIT 1, LOSS OF RECOVERY WITH SI 2 UNIT 1, LOSS OF RECOVERY WITH SI POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 45 of 96 I ____ A_c_T_I_o_N_I_E_x_P_E_c_T_E_o_RE_s_P_o_N_s_E
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__ o _____ _.l Al A2 ATTACHMENT A (Page 1 of 5) G-01 LOCAL MANUAL START Check Green POWER ON Light -LIT
* Panel C-64A
* Panel C-34 Check Overspeed Trip Alarms -CLEAR
* Panel C-64A
* Panel C-34 IF green light is~ lit, THEN transfer control power to alternate source: a. At PAB 8' elevation South of Unit 2 charging pumps, direct PAB operator to shut switch 031-01. b. At C-78, shift to alternate power by swapping paired breakers:
* For annunciators, open breaker 1 and close breaker 2.
* For start circuit 1, open breaker 3 and close breaker 4.
* For control power, open breaker 5 and close breaker 6.
* For field flash, open breaker 7 and close breakers:
Reset mechanical overspeed trip and alarms as follows: a. At C-64, place 43/G-01-ESS, G-01 EOG Engine Start Selector Switch in LOCAL START. b. Reset mechanical overspeed trip. c. At C-64, place 43/G-01-ESS, G-01 EOG Engine Start Selector Switch in AUTO START.
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_____ ...,I ATTACHMENT A (Page 2 of 5) G-01 LOCAL MANUAL START A3 Emergency Start G-01: A4 a. At C-34A, place local/remote transfer switches to LOCAL:
* 43/G-01-1, G-01 EOG Transfer Switch
* 43/G-01-2, G-01 EOG Transfer Switch b. At C-34A, start G-01 by depressing EMERGENCY START pushbutton At C-64, Check G-01 Speed -GREATER THAN OR EQUAL TO 900 RPM b. IF G-01 will NOT emergency start, THEN manually start G-01: 1) At C-64, place 43/G-01-ESS, G-01 EOG Engine Start Selector Switch in LOCAL START. 2) At C-64, depress and hold ENGINE START pushbutton until engine speed rises to idle. 3) At C-64, raise engine speed to 900 rpm by depressing IDLE RELEASE pushbutton.
: 4) IF G-01 CANNOT be started, THEN do not continue and inform Control Room of G-01 status. Perform the following:
: a. At C-64, place 43/G-01-GOV, G-01 Governor Mode Selector Switch to HYO. b. IF diesel speed NOT greater than or equal to 900 RPM, THEN at C-64, raise speed using hydraulic governor control switch.
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_____ __.l A5 ATTACHMENT A (Page 3 of 5) G-01 LOCAL MANUAL START CAUTION Rubber gloves with leather protectors (located in VSG Room cubicle 39) and personnel safety equipment are required to locally operate FFC relay. Contact Control Room To Check G-01 Frequency
-BETWEEN 59.7 Hz AND 60.3 Hz Perform the following:
: a. IF field is NOT flashed, THEN in C-34, manually actuate relay FFC for 3 seconds by pushing yellow tab on relay actuator located behind terminal connections.
: b. At C-64, ensure 43/G-01-GOV, GOl EOG Mode Selector Switch in HYO, THEN adjust frequency using the hydraulic governor control switch. c. IF hydraulic governor control ;;itch NOT functional, THEN adjust frequency using SPEED control knob on faceplate of Woodward governor.
: d. IF frequency CANNOT be maintained, THEN locally shutdown G-01: 1) At C-64, push both engine stop pushbuttons.
: 2) Pull fuel supply cut-off valve operator.
: 3) At C-34A, place output breakers in pull out:
* 1A52-60
* 2A52-73 4) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 48 of 96 I STEP 11 ACTION/EXPECTED RESPONSE 11 RESPONSE NOT OBTAINED I A6 ATTACHMENT A (Page 4 of 5) G-01 LOCAL MANUAL START . Contact Cont~ol Room To Check G-01 Voltage -BETWEEN 4050 Vac AND 4300 Vac Perform the following:
: a. IF field is NOT flashed, THEN in C-34, manually actuate relay FFC for 3 seconds by pushing yellow tab on relay actuator located behind terminal connections.
: b. Contact Control Room to maintain voltage between 4050 Vac and 4300 Vac by adjusting diesel loading. c. IF voltage CANNOT be maintained, THEN locally shutdown G-01: 1) At C-64, push both engine stop pushbuttons.
: 2) Pull fuel supply cut-off valve operator.
: 3) At C-34A, place output breakers in pull out:
* 1A52-60
* 2A52-73 4) Return to procedure and step in effect.
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_____ _.I ATTACHMENT A (Page 5 of 5) G-01 LOCAL MANUAL START A7 Energize Bus lA-05 From Normal Diesel Supply G-01: AB a. Check G-01 -RUNNING b. In VSG room, ensure lA-03 to lA-05 bus tie breaker -OPEN
* 1A52-57 c. At C-35A, ensure G-02 to lA-05 bus tie breaker control switch -OPEN AND IN PULL OUT
* 1A52-66 d. At C-34A, ensure G-01 to bus lA-05 breaker control switch -IN AUTO
* 1A52-60 e. At C-34A, check G-01 to bus lA-05 breaker -CLOSED
* 1A52-60 Check Bus lB-03 -ENERGIZED A9 Return To Procedure And Step In Effect -END-a. Return to procedure and step in effect. b. Return to procedure and step in effect. c. Return to procedure and step in effect. d. Place G-01 to bus lA-05 breaker control switch in AUTO. e. Locally perform the following:
: 1) Try to auto-close breaker by placing control switch to trip position then release. 2) IF breaker will NOT auto-close, THEN manually close breaker control switch. IF bus lA-05 energized, THEN energize bus lB-03: a. Close bus lA-05 feed to lX-13:
* 1A52-58, train A b. Close bus lB-03 normal feed:
* 1B52-l6B, train A POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 50 of 96 E:J I.__ ___ A_c_T_I_o_N_IE_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
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__ N_o_T_o_B_T_A_I_NE_D
_______ I ATTACHMENT B (Page 1 of 5) G-02 LOCAL MANUAL START Bl Dispatch Operator With Key Number 43 To G-02 B2 B3 Check Green POWER ON Light -LIT
* Panel C-65A
* Panel C-35 Check Overspeed Trip Alarms -CLEAR a Panel C-65A
* Panel C-35 IF green light is NOT lit, THEN transfer control power to alternate source: a. At PAB 8' elevation South of Unit 2 charging pumps, direct PAB watch to shut switch 041-01. b. At C-79, shift to alternate power by swapping paired breakers:
* For annunciators, open breaker 1 and close breaker 2.
* For start circuit 1, open breaker 3 and close breaker 4.
* For control power, open breaker 5 and close breaker 6.
* For field flash, open breaker 7 and close breaker 8. Reset mechanical overspeed trip and alarms as follows: a. At C-65, place 43/G-02-ESS, G-02 EOG Engine Start Selector Switch in LOCAL START. b. Reset mechanical overspeed trip. c. At C-65, place 43/G-02-ESS, G-02 EOG Engine Start Selector Switch in AUTO START.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 51 of 96 lsTEPI l._~~-A_c_T_I_o_N_I_E_x~P-E_c_T_E_D~RE~s_P_o_N~S-E~~----''l._~~~~~RE~s-P_o_N~S-E~N-o_T~o-B_T_A~INE~-D~~~~~_.I ATTACHMENT B (Page 2 of 5) G-02 LOCAL MANUAL START B4 Emergency Start G-02: BS a. At C-35A, place local/remote transfer switches to LOCAL:
* 43/G-02-1, G-02 EOG Transfer Switch
* 43/G-02-2, G-02 EOG Transfer Switch b. At C-35A, start G-02 by depressing EMERGENCY START pushbutton At C-65, Check G-02 Speed -GREATER THAN OR EQUAL TO 900 RPM b. IF G-02 will NOT emergency start, THEN manually start G-02: 1) At C-65, place 43/G-02-ESS, G-02 EOG Engine Start Selector Switch in LOCAL START. 2) At C-65, depress and hold ENGINE START pushbutton until engine speed rises to idle. 3) At C-65, raise engine speed to 900 rpm by depressing IDLE RELEASE pushbutton.
: 4) IF G-02 CANNOT be started, THEN do not continue and inform Control Room of G-02 status. Perform the following:
: a. At C-65, place 43/G-02-GOV, G-02 Governor Mode Selector Switch to HYO. b. IF diesel speed NOT greater than ;; equal to 900 RPM, THEN at C-65, raise speed using hydraulic governor control switch.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 52 of 96 LOSS OF ALL AC POWER I STEP 11 ACTION/EXPECTED RESPONSE I [ RESPONSE NOT OBTAINED I B6 ATTACHMENT B (Page 3 of 5) G-02 LOCAL MANUAL START CAUTION Rubber gloves with leather protectors (located in VSG Room cubicle 39) and personnel safety equipment are required to locally operate FFC relay. Check G-02 Frequency
-BETWEEN 59.7 Hz AND 60.3 Hz Perform the following:
: a. IF field is NOT flashed, THEN in C-35, manually actuate relay FFC for 3 seconds by pushing yellow tab on relay actuator located behind terminal connections.
: b. At C-65, ensure 43/G-02-GOV, G02 EDG Mode Selector Switch in HYD, THEN adjust frequency using the hydraulic governor control switch. c. IF hydraulic governor control switch NOT functional, THEN adjust frequency using SPEED control knob on faceplate of Woodward governor.
: d. IF frequency CANNOT be maintained, THEN locally shutdown G-02: 1) Push both engine stop pushbuttons.
: 2) Pull fuel supply cut-off valve operator.
: 3) At C-35A, place output breakers in pull out:
* 1A52-66
* 2A52-67 4) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 53 of 96 lsTEPI l.__~~-A_c_T_I_o_N_I_E_x~P-E_c_T_E_D~RE~s-P_o_N~S-E~~--'ll.__~~~~~RE~s-P_o_N~s-E~N-o_T~o-B_T_A_I_NE~-D~~~~~
..... 1 ATTACHMENT B B7 (Page 4 of 5) G-02 LOCAL MANUAL START Check G-02 Voltage -BETWEEN 4050 Vac AND 4300 Vac Perform the following:
: a. IF field is NOT flashed, THEN in C-35, manually actuate relay FFC for 3 seconds by pushing yellow tab on relay actuator located behind terminal connections.
: b. Contact Control Room to maintain voltage between 4050 Vac and 4300 Vac by adjusting diesel loading. c. IF voltage CANNOT be maintained, THEN locally shutdown G-02: 1) Push both engine stop pushbuttons.
: 2) Pull fuel supply cut-off operator.
valve 3) At C-35A, place output breakers in pull out:
* 1A52-66
* 2A52-67 4) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 54 of 96 E:J I..._ ___ A_c_T_I_o_N_f_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
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__ N_o_T_o_B_T_A_I_NE_D
_____ _.I ATTACHMENT B (Page 5 of 5) G-02 LOCAL MANUAL START BB Energize Bus lA-05 From Alternate Supply G-02: B9 a. Check G-02 -RUNNING b. In VSG Room, ensure lA-03 to lA-05 bus tie breaker -OPEN
* 1A52-57 c. At C-34A, ensure G-01 to lA-05 bus tie breaker control switch -OPEN AND IN PULL OUT
* 1A52-60 d. At C-35A, unlock and place G-02 to bus lA-05 breaker control switch -AUTO
* 1A52-66 e. At C-35A, check G-02 to bus lA-05 breaker -CLOSED
* 1A52-66 Check Bus lB-03 -ENERGIZED BlO Return To Procedure And Step In Effect -END-a. Return to 12rocedure and ste12 in effect. b. Return to 12rocedure and ste12 in effect. c. Return to procedure and ste12 in effect. e. Locally perform the following:
: 1) Try to auto-close breaker by placing control switch to trip position then release. 2) IF breaker will NOT auto-close, THEN manually close breaker control switch. IF bus lA-05 energized, THEN energize bus lB-03: a. Close bus lA-05 feed to lX-13:
* 1A52-58, train A b. Close bus lB-03 normal feed:
* 1B52-16B, train A POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 55 of 96 LOSS OF ALL AC POWER I..__ __ A_c_T_I_o_N_f_E_x_P_E_c_T_E_o_RE_s_P_o_N_s_E
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_____ _.l Cl C2 ATTACHMENT C (Page 1 of 6) G-03 LOCAL MANUAL START CAUTION Flashing fields on both G-03 and G-04 at the same time will overload the 125 VDC power supply. At D-28, Check D-28 125 VDC Control Power -GREATER THAN 115 Vdc
* BUS D-28 Voltage Selector Switch selected to D-28 At Panel C-101, Check Overspeed Trip Alarm -CLEAR Switch to alternate 125 voe control power: a.,!,! power to D-40 NOT available, THEN return to procedure and step in effect. b. At D-40, place fused disconnect to ON:
* 072-40-13
: c. At D-28, place main power transfer switch to OFF:
* 072-28-M d. At D-28, place alternate power transfer switch to ON:
* 072-28-A e. At D-28, check voltage -GREATER THAN 115 Vdc. f. IF 125 voe control power is NOT available, THEN return to procedure and step in effect. Reset overspeed trip and alarm: a. Pull down on gray reset lever to the latched position.
: b. At e-101, depress ALARM RESET pushbutton.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 56 of 96 lsTEPI l._~~-A_c_T_I_o_N_I_E_x~P-E_c_T_E_D~RE~s_P_o_N_s_E~~~~'l._~~~~~RE~s-P_o_N_s_E~-N-o_T~o-B_T_A_I_NE~D~~~~~---'
ATTACHMENT C (Page 2 of 6) G-03 LOCAL MANUAL START C3 Auto Start G-03: a. At C-81, place 43/G-03-LRS, G-03 EDG Local/Remote Start Selector Switch to LOCAL b. At C-81, ensure 43/G-03-GOV, G-03 EDG Mode Selector Switch in AUTO c. At C-81, depress SHUTDOWN RESET pushbutton
: d. Check G-03 -RUNNING C4 At C-81, Depress ALARM RESET Pushbutton
: d. Fast start G-03: 1) At C-81, depress FAST START pushbutton.
: 2) IF G-03 did NOT fast start, manually start G-03: a) At C-81, depress SHUTDOWN RESET pushbutton.
b) At C-81, depress VOLTAGE SHUTDOWN RESET pushbutton.
c) At C-81, depress ALARM RESET pushbutton.
d) At C-81, place 43/G-03-GOV G-03 EOG GOVERNOR MODE SELECTOR SWITCH, to HYD. e) IF G-03 is NOT running, THEN depress FAST START pushbutton.
: 3) IF G-03 CANNOT be started,~
return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 E'age 57 of 96 I STEP 11 ACTION/EXPECTED RESPONSE 11 RESPONSE NOT OBTAINED I ....___ ____ ____.L.....-
_____ _____. cs ATTACHMENT C (Page 3 of 6) G-03 LOCAL MANUAL START At C-81, Check G-03 Frequency
-BETWEEN 59.7 Hz AND 60.3 Hz Perform the following:
: a. IF field is NOT flashed, THEN at C-81, depress VOLTAGE SHUTDOWN RESET pushbutton.
: b. At C-81, ensure 43/G-03-GOV, G03 EOG Mode Selector Switch in HYO, THEN adjust frequency using the hydraulic governor control switch. c. IF hydraulic governor control ;;itch NOT functional, THEN adjust frequency using SPEED control knob on faceplate of Woodward governor.
: d. IF frequency CANNOT be maintained, THEN locally shutdown G-03: 1) At C-81, push both engine stop pushbuttons:
* Black engine stop
* Red emergency stop 2) At C-81, place output breakers in pull out:
* 1A52-80
* 2A52-87 3) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 58 of 96 lsTEPI l._~~-A_c_T_I_o_N_I_E_x_P_E_c_T_E_D~RE~s_P_o_N_s_E~~~ll
..... ~~~~-RE~s-P_o_N~S-E~N-o_T~o-B_T_A_I_NE~D~~~~~
.... 1 ATTACHMENT C C6 (Page 4 of 6) G-03 LOCAL MANUAL START At C-81, Check G-03 Voltage -BETWEEN 4050 Vac AND 4300 Vac
* AC Voltmeter
-Generator Perform the following:
: a. IF field is NOT flashed, THEN at C-81, depress VOLTAGE SHUTDOWN RESET pushbutton.
: b. IF voltage CANNOT be maintained, THEN locally shutdown G-03: 1) At C-81, push both engine stop pushbuttons:
* Black engine stop
* Red emergency stop 2) At C-81, place output breakers in pull out:
* 1A52-80
* 2A52-87 3) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 59 of 96 I ___ A_c_T_I_o_N_I_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
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_____ -1 ATTACHMENT C (Page 5 of 6) G-03 LOCAL MANUAL START C7 Energize Bus lA-06 From Normal Diesel Supply G-03: a. Check G-03 -RUNNING b. In DGB room, ensure lA-04 to lA-06 bus tie breaker -OPEN
* 1A52-77 c. At C-82, ensure G-04 to lA-06 bus tie breaker control switch -OPEN AND IN PULL OUT
* 1AS2-86 d. At C-81, ensure G-03 to bus lA-06 breaker control switch -IN AUTO
* 1A52-80 e. At C-81, check G-03 to bus lA-06 breaker -CLOSED
* 1A52-80 a. Return to procedure and step in effect. b. Return to 12rocedure and ste12 in effect. c. Return to procedure and ste12 in effect. d. Place G-03 to bus lA-06 breaker control switch in AUTO. e. Locally perform the following:
: 1) Try to auto-close breaker by placing control switch to trip position then release. 2) IF breaker will NOT auto-close, THEN perform the following:
a) Turn on synchronizing switch:
* G-03 Diesel Generator to bus lA-06 Synchroscope b) Manually close breaker control switch:
* 1A52-80 c) Turn off synchronizing switch:
* G-03 Diesel Generator to bus lA-06 Synchroscope POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 60 of 96 E:J I...._ ___ A_c_T_I_o_N_I_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
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__ N_o_T_o_B_T_A_I_NE
__ D _____ .....,I cs ATTACHMENT C (Page 6 of 6) G-03 LOCAL MANUAL START Check Bus lB-04 -ENERGIZED IF lA-06 energized, THEN energize bus lB-04: a. Close bus lA-06 feed to lX-14:
* 1A52-84, train B b. Close bus lB-04 normal feed:
* 1B52-17B, train B C9 Return To Procedure and step in effect -END-POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 61 of 96 lsTEPI l._~~-A_c_T_I_o_N_f_E~x_P_E_c_T_E_D~RE~s_P_o_N~S-E~~---'ll.__~~~~~RE--s_P_o_N_s_E
___ N_o_T __ o_B_T_A_I_NE
___ D ________ ..... 1 ATTACHMENT D (Page 1 of 6) G-04 LOCAL MANUAL START Dl Dispatch Operator With Key Number 43 To G-04 D2 D3 CAUTION Flashing fields on both G-03 and G-04 at the same time will overload the 125 VDC power supply. At D-40, Check D-40 125 VDC Control Power -GREATER THAN 115 Vdc
* BUS D-40 Voltage Selector Switch selected to D-40 At Panel C-102, Check Overspeed Trip Alarm -CLEAR Switch to alternate 125 voe control power: a. IF power to D-28 NOT available, THEN return to procedure and step in effect. b. At D-28, place fused disconnect to ON:
* 072-28-13
: c. At D-40, place main power transfer switch to OFF:
* 072-40-M d. At D-40, place alternate power transfer switch to ON:
* 072-40-A e. At D-40, check voltage -GREATER THAN 115 Vdc. f. IF 125 voe control power is NOT available, THEN return to procedure and step in effect. Reset overspeed trip and alarm: a. Pull down on gray reset lever to the latched position.
: b. At e-102, depress ALARM RESET pushbutton.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 62 of 96 I STEP 11 ACTION/EXPECTED RESPONSE 11 RESPONSE NOT OBTAINED I ATTACHMENT D (Page 2 of 6) G-04 LOCAL MANUAL START D4 Auto Start G-04: a. At C-82, place 43/G-04-LRS, G-04 EOG Local/Remote Start Selector Switch, to LOCAL b. Ensure 43/G-04-GOV, G-04 EOG Governor Mode Selector Switch, in AUTO c. At C-82, depress SHUTDOWN RESET pushbutton
: d. Check G-04 -RUNNING D5 At C-82, Depress ALARM RESET Pushbutton
: d. Fast start G-04: 1) At C-82, depress FAST START pushbutton.
: 2) IF G-04 did NOT fast start, THEN manually start G-04: a. At C-82, depress SHUTDOWN RESET pushbutton.
: b. At C-82, depress VOLTAGE SHUTDOWN RESET pushbutton.
: c. At C-82, depress ALARM RESET pushbutton.
: d. At C-82, place 43/G-04-GOV, G-04 EOG Governor Mode Selector Switch, to HYO. e. IF G-04 is NOT running, THEN depress FAST START pushbutton.
: 3) IF G-04 CANNOT be started, THEN return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 63 of 96 lsTEPI 1 ...... ~~-A_c_T_I_o_N_f_E_x~P-E_c_T_E_n~RE~s-P_o_N~s_E~~--'11
....... ~~~~~RE~s-P_o_N~s-E~N-o_T~o-B_T_A~INE~-n~~~~~-''
ATTACHMENT D D6 (Page 3 of 6) G-04 LOCAL MANUAL START At C-82, Check G-04 Frequency
-BETWEEN 59. 7 HZ AND 60. 3 HZ Perform the following:
: a. IF field is NOT flashed, THEN at C-82, depress VOLTAGE SHUTDOWN RESET pushbutton.
: b. At C-82, ensure 43/G-04-GOV, G04 EDG Mode Selector Switch in HYD, THEN adjust frequency using the hydraulic governor control switch. c. IF hydraulic governor control ;;itch NOT functional, THEN adjust frequency using SPEED control knob on faceplate of Woodward governor.
: d. IF frequency CANNOT be maintained, THEN locally shutdown G-04: 1) At C-82, push both engine stop pushbuttons:
* Black engine stop
* Red emergency stop 2) At C-82, place output breakers in pull out:
* 1A52-86
* 2A52-93 3) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 64 of 96 EJ I.__ __ A_c_T_I_o_N_f_E_x_P_E_c_T_E_n_RE_s_P_o_N_s_E
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_____ .... 1 D7 ATTACHMENT D (Page 4 of 6) G-04 LOCAL MANUAL START At C-82, Check G-04 Voltage -BETWEEN 4050 Vac AND 4300 Vac
* AC Voltmeter
-Generator Perform the following:
: a. IF field is NOT flashed, THEN at C-82, depress VOLTAGE SHUTDOWN RESET pushbutton.
: b. IF voltage CANNOT be maintained, THEN locally shutdown G-03: 1) At C-82, push both engine stop pushbuttons:
* Black engine stop
* Red emergency stop 2) At C-82, place output breakers in pull out:
* 1A52-86
* 2A52-93 3) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 65 of 96 lsTEPI l...._~~A~c_T_I_o_N_/_E_x_P_E_c_T_E_D~RE~s_P_o_N_s_E~~__.ll.._~~~~-RE~-s-P_oN~S-E~N-o_T~o-B_T_A_I_NE~D~~~~~_.I ATTACHMENT D (Page 5 of 6) G-04 LOCAL MANUAL START DB Energize Bus lA-06 From Alternate Supply G-04: a. Check G-04 -RUNNING b. In DGB room, ensure lA-04 to lA-06 bus tie breaker -OPEN
* 1A52-77 c. At C-81, ensure G-03 to lA-06 bus tie breaker control switch -OPEN AND IN PULL OUT
* 1A52-80 d. At C-82 unlock and place G-04 to bus lA-06 breaker control switch in AUTO
* 1A52-86 e. At C-82, check G-04 to bus lA-06 breaker -CLOSED
* 1A52-86 a. Return to 12rocedure and ste12 in effect. b. Return to 12rocedure and ste12 in effect. c. Return to 12rocedure and ste12 in effect. e. Locally perform the following:
: 1) Try to auto-close breaker by placing control switch to trip position then release. 2) IF breaker will NOT auto-close, THEN perform the following:
a) Turn on synchronizing switch:
* G-04 Diesel Generator to bus lA-06 Synchroscope b) Manually close breaker control switch:
* 1A52-86 c) Turn off synchronizing switch:
* G-04 Diesel Generator to bus lA-06 Synchroscope POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFE TY RELATED Revision 69 Page 66 of 96 lsTEPI l.__~~A~c_T_I_o_N_I_E_x_P_E_c_T_E_D~RE~s_P_o_N_s_E~~---'" ..... ~~~~~RE~s_P_o_N_s_E~N-o_T~o-B_T_A_I_NE~D~~~~~_,'
ATTACHMENT D (Page 6 of 6) G-04 LOCAL MANUAL START D9 Check Bus lB-04 -ENERGIZED D10 Return To Procedure and step in effect -END-IF lA-06 energized, THEN energize bus lB-04: a. Close bus lA-06 feed to lX-14:
* 1A52-84, train B b. Close bus lB-04 normal feed:
* 1B52-17B, train B POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 67 of 96 LOSS OF ALL AC POWER I ___ A_c_T_I_o_N_f_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
__ __.I I.__ _____ RE __ s_P_o_N_s_E_N_o_T_o_B_T_A_I_NE_D
_____ _.I ATTACHMENT E (Page 1 of 6) POWER RESTORATION USING GAS TURBINE El Start G-05 Gas Turbine E2 NOTE When G-501, Auxiliary Diesel Generator is running, the G-501 fuel tank MUST be filled by an operator at 45 minute intervals.
The estimated maximum G-501 run time is one hour. a. Check G-501 GT GENERATOR AUXILIARY DG RUNNING light -LIT b. Check G-05 REMOTE STATION OPERATIVE light -LIT c. Check READY TO START light -LIT d. Place GENERATOR RATE SELECTOR switch to NORMAL e. Depress START pushbutton
: f. Depress MINIMUM LOAD pushbutton Check G-05 GT GENERATOR READY TO SYNCHRONIZE Light -LIT a. At G-501 Engine Control Panel, place Auto-Off-Start selector switch to Start position b. Locally, Align G-501 output to supply G-05 Auxiliaries by performing the following:
: 1) Locally at B-507, place 43/52-N-52-E G-05 Auxiliaries Auto/Manual Transfer Switch to MAN. 2) Locally at 52-T, Emergency Pwr to B-502 TSC, place 52T switch to OPEN. 3) Locally at B-507, place 52-E Erner Pwr to G-05 GT Gen/Auxiliaries Control Switch to CLOSE. c. Align GT Auxiliaries per OI 110, GAS TURBINE OPERATION
: e. Locally start G-05 per OI 110, GAS TURBINE OPERATION DO NOT CONTINUE until G-05 READY TO SYNCHRONIZE light is lit.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 68 of 96 LOSS OF ALL AC POWER I STEP 11 ACTION/EXPECTED RESPONSE 11 RESPONSE NOT OBTAINED I .__ __________________
__..._ ______________________
__. E3 E4 ES E6 ATTACHMENT E (Page 2 of 6) POWER RESTORATION USING GAS TURBINE Ensure POWER TO/FROM H-01 BUS Breaker -CLOSED
* H52-10 Ensure The Following Electrical Breakers -OPEN
* 1A52-01, bus lA-01 Normal Feed
* 1A52-36, bus lA-03 Normal Feed
* 1A52-37, lA-03 to lA-01 Bus Tie Breaker
* 1A52-40, lA-03 to 2A-03 Bus Tie Breaker
* 1A52-5 6, bus lA-04 Normal Feed
* 1A52-55, lA-04 to lA-02 Bus Tie Breaker
* 1A52-17, bus lA-02 Normal Feed
* 1A52-52, lA-04 to 2A-04 Bus Breaker IF BOTH units have entered ECA-0.0, THEN isolate the 13.8 Bus from switchyard by ensuring the following breakers open:
* H52-20
* H52-30 Energize Bus H-01 And Bus H-02 From G-05: Tie kV a. Turn on G-05 GT GENERATOR MAIN BREAKER SYNCHROSCOPE switch b. Close G-05 main breaker:
* H52-G05 c. Turn on H-02 TO H-01 BUS TIE SYNCHROSCOPE switch d. Close H-02 to H-01 bus tie breaker:
* H52-21 Ensure the following breakers are open:
* H52-20
* H52-31
* H52-21 IF breakers will NOT close, THEN perform the following:
: 1. Check status of the following lockouts:
* lX-04 lockout
* H-01 bus lockout
* H-02 bus lockout 2. If lockouts are NOT tripped, THEN locally shut stored energy breakers
* H52-G05
* H52-21 POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 69 of 96 I STEP I.__ I ___ A_c_T_I_o_N_I_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
__ ___.I I.__ _____ RE __ s_P_o_N_s_E_N_o_T_o_B_T_A_I_NE_D
_____ ....,I ATTACHMENT E (Page 3 of 6) POWER RESTORATION USING GAS TURBINE E7 Continue Operation Of Gas Turbine Per OI-110, GAS TURBINE OPERATION ES Check Bus H-02 -ENERGIZED NOTE IF H-02 CANNOT be energized, THEN perform the following:
: a. Evaluate the plant status to determine if the gas turbine is required to support Appendix R or other plant equipment.
: b. Shut down gas turbine if desired. c. Operate G-501 to power the TSC per OI-35, ELECTRICAL EQUIPMENT OPERATION.
: d. Return to procedure and step in effect. Synchroscope operation required for lA-03 / lA-04 feed breaker closure. E9 Restore Power To Bus lA-03 And Bus lA-04: ElO a. Reset and close bus H-02 feed to lX-04:
* H52-22 b. Reset and close bus lA-03 normal feed:
* 1A52-36 c. Reset and close bus lA-04 normal feed:
* 1A52-56 Check Bus lA-03 And Bus lA-04 -AT LEAST ONE ENERGIZED Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 70 of 96 lsTEPI I ACTION/EXPECTED RESPONSE II RESPONSE NOT OBTAINED I .._ _________________
__,.._ _____________________
_. ATTACHMENT E (Page 4 of 6) POWER RESTORATION USING GAS TURBINE Ell Energize Bus lA-05 From lA-03: a. Check bus lA-03 -ENERGIZED
: b. Ensure G-01 to bus lA-05 breaker -OPEN
* 1A52-60 c. Ensure G-02 to bus lA-05 breaker -OPEN
* 1A52-66 d. Turn on synchronizing switch for lA-03 to lA-05 bus tie breaker:
* 1A52-57 e. Trip and close lA-03 to lA-05 bus tie breaker:
* 1A52-57 a. Go to Step El2.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 71 of 96 jsTEPI 'L-~~-A_c_T_I_o_N~/E~x_P_E_c_T_E_D~RE~s_P_o_N~s_E~~__,111--~~~~~RE~s-P_o_N_s_E~-N-o_T~O-B_T_A_I_NE~-D~~~~~-'I ATTACHMENT E (Page 5 of 6) POWER RESTORATION USING GAS TURBINE E12 Energize Bus lA-06 From lA-04: E13 a. Check bus lA-04 -ENERGIZED
: b. Ensure G-03 to bus lA-06 breaker -OPEN
* 1A52-80 C. Ensure G-04 to bus lA-06 breaker -OPEN
* 1A52-86 d. Trip and close lA-04 normal feed to lA-06:
* 1A52-54 e. Turn on synchronizing switch for lA-04 to lA-06 bus tie breaker:
* 1A52-77 f. Trip and close lA-04 to lA-06 bus tie breaker:
* 1A52-77 Check Bus lA-05 And Bus lA-06 -AT LEAST ONE ENERGIZED
: a. Go to Step E13. Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 72 of 96 lsTEPI l._~~A~c_T_I_o_N_f_E_x_P_E_c_T_E_D~RE~s_P_o_N_s_E~~___.11
..... ~~~~-RE~-s_P_o_N_s_E~N_o_T~o_B_T_A_I_NE~D~~~~~
.... 1 ATTACHMENT E E14 (Page 6 of 6) POWER RESTORATION USING GAS TURBINE Check 480 Vac Safeguards Buses -AT LEAST ONE ENERGIZED
* lB-03, train A
* lB-04, train B Try to restore power to either bus as follows: a. IF bus lA-05 energized, THEN energize bus lB-03: 1) Close bus lA-05 feed to lX-13:
* 1A52-58, train A 2) Close bus lB-03 normal feed:
* 1B52-16B, train A b. IF lA-06 energized, THEN energize bus lB-04: 1) Close bus lA-06 feed to lX-14:
* 1A52-84, train B 2) Close bus lB-04 normal feed:
* 1B52-17B, train B E15 Return To Step 46. -END-POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 73 of 96 I STEP 11 ACTION/EXPECTED RESPONSE 11 RESPONSE NOT OBTAINED I .__ _________________
__,.._ _____________________
.... Fl ATTACHMENT F (Page 1 of 3) BACKFEED TO 480 VAC SAFEGUARDS BUSES Check lA-01 And lA-02 -At LEAST ONE ENERGIZED Perform the following:
: a. Energize lA-01 and lA-02 per AOP-18, ELECTRICAL SYSTEM MALFUNCTION.
F2 Energize Bus lB-01 From Bus lA-01 And Bus lB-02 From Bus lA-02: a. Close bus lA-01 feed to lX-11:
* 1A52-02, train A b. Close bus lB-01 normal feed:
* 1B52-04B, train A c. Close bus lA-02 feed to lX-12:
* 1A52-15, train B d. Close bus lB-02 normal feed:
* 1B52-05B, train B POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 74 of 96 jsTEPI j~~~-A_c_T_I_o_N_I_E_x_P_E_c_T_E_D~RE~s_P_o_N_s_E~~----'1
..... ~~~~-RE~-s-P_oN~S-E~N-o_T~o-B_T_A_I_NE~D~~~~~
.... 1 ATTACHMENT F (Page 2 of 3) BACKFEED TO 480 VAC SAFEGUARDS BUSES NOTE The breaker trip for transformers lX-11 and lX-12 is 180 amps each. Refer to AOP-22 UNIT 1, EDG LOAD MANAGEMENT, for equipment load ratings. F3 Backfeed Bus lB-03 From Bus lB-01 And Bus lB-04 From Bus lB-02: a. Place Unit 1 service water pumps in pull out:
* P-32A, train A
* P-32B, train A
* P-32C, train B b. Place DC control power fuse block for each tie breaker to OFF: 1) In lB-03, place DC control power fuse block for lB-01 to lB-03 bus tie breaker to OFF:
* 1B52-15C, train A 2) In lB-04, place DC control power fuse block for lB-04 to* lB-02 bus tie breaker to OFF:
* 1B52-18C, train B c. Ensure normal feed breakers open:
* 1B52-16B for lB-03
* 1B52-17B for lB-04 d. Locally close tie breakers:
* 1B52-l5C, lB-03 to lB-01 bus tie breaker
* 1B52-18C, lB-04 to lB-02 bus tie breaker POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 75 of 96 lsTEPI l~~~A~c_T_I_o_N_I_E_x_P_E_c_T_E_D~RE~s_P_o_N_s_E~~~ll
...... ~~~~-RE~S~Po_N~S-E~N-o_T~o-B_T_A_I_NE~D~~~~~
.... 1 ATTACHMENT F (Page 3 of 3) BACKFEED TO 480 VAC SAFEGUARDS BUSES F4 Return To Procedure and step in effect -END-POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ATTACHMENT G (Page 1 of 3) ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 76 of 96 ALIGNING EQUIPMENT TO ALTERNATE POWER SOURCE Gl IF power is NOT established to H-01, THEN return to procedure and step in effect. G2 IF placing charging pump 1P-2A in operation from alternate power supply, THEN perform the following:
: a. Locally ensure 1P-2A normal feeder breaker open:
* 1B52-13A b. Outside charging pump cubicles, align 1P-2A charging pump normal/alternate transfer switch to alternate power source:
* At 1B313A-B854B, place transfer switch in A2 ON -A4 OFF position.
: c. Place 1P-2A charging pump controller in MANUAL at minimum speed:
* 1HC-428A d. At C-45, Alternate Shutdown Control Panel, close 1P-2A alternate feeder breaker:
* B52-54B e. At 1C04, place 1P-2A charging pump switch to START. f. IF the pump fails to start, THEN wait for the VFD to initialize and then place the control switch to start. g. Adjust charging line flow controller to full open:
* lHC-142 h. At 1C04, place "In Alternate Control" placard next to 1P-2A charging pump control switch.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ATTACHMENT G (Page 2 of 3) ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 77 of 96 ALIGNING EQUIPMENT TO ALTERNATE POWER SOURCE CAUTION If an undervoltage occurs on B-08 or B-09, local manual restart of associated service water pump is required.
G3 placing a service water pump in operation from bus B-08, THEN perform the following:
: a. Select one service water pump to be powered from alternate supply: o P-32B o P-32F b. Locally ensure normal feeder breaker for selected service water pump open: o 1B52-l1C for P-32B o 2B52-34B for P-32F c. In Room G-01, place alternate power selector switch to desired service water pump:
* B8540 d. In Room G-01, place normal/alternate transfer switch for selected service water pump to B-08 power supply: o 1B3l1C-B854D for P-32B o 2B334B-B854D for P-32F e. At C-45, Alternate Shutdown Control Panel, check selected service water pump switches properly aligned to B-08 power supply using indicating lights. f. At C-45, close P-32B/F alternate feeder breaker:
* B52-54D g. At COl, place "In Alternate Control" placard next to service water pump control switch.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ATTACHMENT G (Page 3 of 3) ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 78 of 96 ALIGNING EQUIPMENT TO ALTERNATE POWER SOURCE G4 IF placing a service water pump in operation from bus B-09, THEN perform the following:
: a. Select one service water pump to be powered from alternate supply: o P-32C o P-32E b. Locally ensure normal feeder breaker for selected service water pump open: o 1B52-20C for P-32C o 2B52-27C for P-32E c. In Room G-02, place alternate power selector switch to desired service water pump:
* B957D d. In Room G-02, place normal/alternate transfer switch for selected service water pump to B-09 power supply: o 1B420C-B957D for P-32C o 2B427C-B957D for P-32E e. At C-45, Alternate Shutdown Control Panel, check selected service water pump switches properly aligned to B-09 power supply using indicating lights. f. At C-45, close P-32C/E alternate feeder breaker:
* B52-57D IF service water pump P-32C/E breaker B52-57D failed to close or remain closed, THEN perform the following:
: 1) At B-09 place the B52-57D breaker charging spring motor control switch to ON. 2) WHEN the breaker closing spring is charged,~
place the B52-57D breaker charging spring motor control switch to OFF. 3) Close service water pump P-32C/E breaker B52-57D. g. At COl, place "In Alternate Control" placard next to service water pump control switch. GS Return to procedure and step in effect. -END-POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ATTACHMENT H (Page 1 of 3) LOCAL SHUTTING OF MSIV NOTE ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 79 of 96 Throughout this attachment, "affected" refers to the MSIV being shut. Hl IF Control Room annunciator CO2 D 4-5, UNIT 1 SAFEGUARDS DC CONTROL POWER FAILURE, is lit, THEN ensure MSIV solenoid power supplies are on:
* 072-16-2, train A
* 072-21-2, train B H2 Shut affected MSIV using local pushbutton:
: a. IF shutting lMS-2018, THEN press both pushbuttons in lRK-33. ---* lMS PB-2018A, train A
* lMS PB-2018B, train B b. IF -shutting lMS-2017, THEN press both pushbuttons in lRK-34.
* lMS PB-2017A, train A
* lMS PB-2017B, train B POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ATTACHMENT H (Page 2 of 3) LOCAL SHUTTING OF MSIV ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page BO of 96 H3 IF affected MSIV will NOT shut using local pushbuttons, THEN vent off air from affected MSIV as follows: a. Obtain two painted red combination wrenches. (MSIV cabinet) b. Shut instrument air supply valve to affected MSIV. o IA-638 for lMS-2018 o IA-636 for lMS-2017 c. Remove end cap downstream of S/G header test isolation valve of affected MSIV. o 1MS-331A for S/G A o 1MS-331B for S/G B d. Vent air off affected MSIV operator by opening S/G header test isolation valve. o 1MS-331A for S/G A o 1MS-331B for S/G B e. Return painted red combination wrenches to MSIV cabinet. f. Red lock combination wrenches.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ATTACHMENT H (Page 3 of 3) LOCAL SHUTTING OF MSIV CAUTION ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 81 of 96 Rapid movement of the ratcheting torque wrench may occur when valve movement starts. NOTE The Direction Position Switch on the ratchet is color coded and unit designated for unit applicability.
H4 IF affected MSIV will NOT shut by venting off air, THEN perform the following:
: a. On 66' PAB near waste distillate tanks, open the M&TE box using the M&TE key and obtain the following equipment:
* One ratcheting torque wrench
* One 2 3/4" socket b. Apply wrench to the nut on the end of the valve shaft and shut affected MSIV. HS Notify Control Room of affected MSIV status. -END-POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 69 LOSS OF ALL AC POWER Page 82 of 96 ATTACHMENT I (Page 1 of 2) CONTAINMENT ISOLATION VALVES PANEL A COMMENT DESCRIPTION TRAIN lCV-1296 Auxiliary charging line A lRC-538 Pressurizer relief tank to gas analyzer A lWG-1788 Reactor coolant drain tank to gas A analyzer lWL-1698 Reactor coolant drain tank to -19 ft A sump 1WL-1003A Reactor coolant drain tank pump suction A 1WL-1003B Reactor coolant drain tank pump suction A lRC-508 Reactor makeup water to containment A or B lRC-539 Pressurizer relief tank to gas analyzer B lWG-1789 Reactor coolant drain tank to gas B analyzer lSI-846 Accumulator nitrogen supply A or B lWL-1 721 Reactor coolant drain tank pumps B suction lWL-1 723 Sump A drain A lSC-951 Pressurizer steam sample A lSC-953 Pressurizer liquid sample A lWL-1728 Sump A drain B 1SC-966A Pressurizer steam sample A or B 1SC-966B Pressurizer liquid sample A or B POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 69 LOSS OF ALL AC POWER Page 83 of 96 ATTACHMENT I (Page 2 of 2) CONTAINMENT ISOLATION VALVES PANEL B COMMENT DESCRIPTION TRAIN lCC-769 Component cooling water outlet from A or B excess letdown heat exchanger lCV-313 Reactor coolant pump seal return A lCV-371 Letdown line A lMS-5958 Steam generator blowdown A or B lMS-5959 Steam generator blowdown A or B lWG-1786 Reactor coolant drain tank vent A 1CV-313A Reactor coolant pump seal return B 1CV-371A Letdown line B lWG-1787 Reactor coolant drain tank vent B 1RM-3200C RE-211/212 supply A 1RM-3200A RE-211/212 return A or B lMS-2083 Steam generator A sample A or B lMS-2084 Steam generator B sample A or B lSC-955 Reactor coolant hot leg sample A lIA-3047 Instrument air line A or B 1RM-3200B RE-211/212 supply B 1SC-966C Reactor coolant hot leg sample A or B lIA-3048 Instrument air line A or B -END-POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 84 of 96 ATTACHMENT J {Page 1 of 1) EQUIPMENT AVAILABLE DURING LOSS OF ALL AC POWER COMMUNICATIONS
* NAWAS
* Outside line in TSC
* ATCo telephone
* Operations radio talkgroups OPS 1, OPS 2, OPS 3 and OPS 4
* Interplant trouble circuit
* Motorola radio to radio channels PNT to PNTl and PNT to PNT 2 * * *
* LIGHTING Station Battery Fixed (EB) DC Lights
* Vital switchgear room
* Diesel room (doorway only)
* Control Room
* Cable spreading room Fixed Emergency DC Sealed Beam Lanterns (EL) Gai-tronics PBX telephone NRC telephone Security radio 1 and Security system talkgroups 2 Security
* These lanterns are located along all entry and egress routes to both safe and alternate shutdown equipment and total approximately 100.
* Each individual lantern battery pack is designed for up to 8 hours of operation allowing for restoration of normal AC power. Portable Emergency Lanterns
* C59
* TSC
* Auxiliary feed tunnel
* Unit 2 non-nuclear room
* Brigade ready
* Fire cart, Unit 2 turbine hall, Elevation 8 ft. -END-POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 85 of 96 RESPONSE NOT OBTAINED I STEP 11 ACTION/EXPECTED RESPONSE I -------------------------------------------1 Kl Verify AFW flow: ATTACHMENT K (Page 1 of 2) ESTABLISHING HEAT SINK a. AFW flow-GREATER THAN OR EQUAL TO 230 gprn a. Establish feed flow from SSG pump as follows: 1) Ensure adequate power is available.
Refer to AOP-22 UNIT 1, EOG LOAD MANAGEMENT, for KW ratings. 2) Place selected Stripping Logic Override Switch to the OVERRIDE position.
* P-38BX-CS
: 3) Start Standby Stearn Generator feed pump:
* P-38B 4) Verify valve alignment:
a) Ensure Unit 1 valve OPEN:
* AF-4021, train B b) Ensure Unit 2 valve SHUT: ** AF-4020 c) Manually align valve as necessary to establish flow greater than or equal to 230 gprn:
* lAF-4019 d) IF feed flow has been established, THEN return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 86 of 96 I STEP IL-I ___ A_c_T_I_o_N_I_E_x_P_E_c_T_E_n_RE_s_P_o_N_s_E
___ ....,.I ______ RE_s_P_o_N_s_E_N-:._o
__ T_o_B_T_A_INE
__ n _____ --1 ATTACHMENT K (Page 2 of 2) ESTABLISHING HEAT SINK -END-IF any feed flow to at least one S/G is verified, THEN perform the following:
: 1) Maintain feed flow to restore S/G level to greater than [51%] 32%. IF feed flow is NOT verified, THEN establish AFW from Unit 2 by performing the following:
: 2) Locally open the following valves:
* lAF-192, Unit 1 AFW Cross-connect Valve
* 2AF-192, Unit 2 AFW Cross-connect Valve 3) Start the Unit 2 motor driven AFW pump and monitor total flow using Unit 2 indications:
* 2P-53 4) Align AFW valve(s) to provide GREATER THAN 230 gpm:
* 1AF-4074A, for S/G A
* 1AF-4074B, for S/G B POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 87 of 96 RESPONSE NOT OBTAINED I STEP 11 ACTION/EXPECTED RESPONSE I '--------------------1---------------------~
ATTACHMENT L (Page 1 of 3) ESTABLISHING LOW PRESSURE FEEDWATER FLOW Ll Initiate FSG-3 UNIT 1, ALTERNATE LOW PRESSURE FEEDWATER.
L2 low pressure feedwater source is ready to provide flow,~ continue with Step L3. NOTE If SI accumulators are isolated or vented and S/G pressures are less than or equal to 320 psig, then S/G pressures and core exit thermocouples should be controlled at less than or equal to existing values. L3 IF shutoff pressure of low pressure feedwater source is greater than 320 psig, THEN perform the following:
: a. Depressurize selected S/G(s) below shutoff pressure of low pressure source using S/G atmospheric steam dumps manually or locally.
* lMS-2016
* lMS-2015 b. IF the source range or intermediate range startup rate is positive, THEN perform the following:
: 1) Control RCS temperature using S/G ADVs to maintain the reactor subcritical.
* lMS-2016
* lMS-2015 2) WHEN S/G ADVs have no little effect on RCS temperature, THEN decrease S/G pressure below shutoff pressure of low pressure source. c. Establish feed flow at target value based on decay heat.
* Reference FSG-3 UNIT 1, ALTERNATE LOW PRESSURE FEEDWATER ATTACHMENT A for valves. d. Maintain S/G pressure at 320 psig using SG ADVs manually or locally:
* lMS-2016
* lMS-2015 e. IF the reactor remains subcritical, THEN control feed flow to maintain S/G narrow range level between [51%] 32% and 63%.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 88 of 96 RESPONSE NOT OBTAINED I STEP 11 ACTION/EXPECTED RESPONSE I '--------------------L.-----------------------1 ATTACHMENT L (Page 2 of 3) ESTABLISHING LOW PRESSURE FEEDWATER FLOW f. IF the source range or intermediate range startup rate is positive, THEN perform the following:
* Control feed flow to increase core exit thermocouples between 428&deg;F and 547&deg;F to prevent criticality.
* WHEN boration or Xenon adds negative reactivity to allow cooldown to 428&deg;F, THEN continue depressurization.
NOTE In order to establish adequate feedwater flow in the following step, it will be necessary to feed and steam S/G's that are close to dry and at low pressure.
L4 IF shutoff pressure of low pressure feedwater source is less than 320 psig, perform the following:
: a. Depressurize selected S/G(s) below shutoff pressure of low pressure source using S/G atmospheric steam dumps manually or locally while maintaining core exit thermocouples between 428&deg;F and 547&deg;F.
* lMS-2016
* lMS-2015 b. IF the source range or intermediate range startup rate is positive, THEN perform the following:
: 1) Control RCS temperature using S/G ADVs to maintain the reactor subcritical.
* lMS-2016
* lMS-2015 2) WHEN S/G ADVs have no little effect on RCS temperature, THEN decrease S/G pressure below shutoff pressure of low pressure source. c. Establish feed flow at target value based on decay heat.
* Reference FSG-3 UNIT 1, ALTERNATE LOW PRESSURE FEEDWATER ATTACHMENT A for valves. d. Lower S/G pressure as required to increase feed flow. e. IF the source range or intermediate range startup rate is positive, THEN perform the following:
* Control feed flow to increase core exit thermocouples between 428&deg;F and 547&deg;F to prevent criticality.
* WHEN boration or Xenon adds negative reactivity to allow cooldown to 428&deg;F, THEN continue depressurization.
: f. Control feed flow and S/G pressures to maintain core exit TCs at 428&deg;F POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 89 of 96 I STEP I I...._ ___ A_c_T_I_o_N_/_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
___ ""I ______ RE_s_P_o_N_s_E_N-:.._o
__ T_o_B_T_A_I_NE_o
_____ ----1 ATTACHMENT L (Page 3 of 3) ESTABLISHING LOW PRESSURE FEEDWATER FLOW LS IF the low pressure source maintains S/G narrow range level between [51%] 32% and 63%, THEN go to Step L9. L6 IF another feedwater source is subsequently established to the same S/G(s), THEN control S/G ADVs steam dump to allow S/G pressure to increase and ~ver S/G level, while avoiding an uncontrolled cooldown.
L7 IF another feedwater source restores a different S/G narrow range level above [51%] 32% and feeding with the low pressure source is no longer necessary to remove decay heat, THEN perform the following:
: a. Isolate the low pressure source from the feed line. b. IF S/G level fed from the low pressure source is less than [51%] 32%, THEN close associated S/G ADV. LS WHEN ECA-0.0 is exited,~ consider the following in subsequent recovery:
* Do not enter CSP-H.l due to low S/G levels and low feedwater flow alone, unless low pressure feed source is not adequate to maintain core exit TCs stable.
* IF main feedwater OR condensate is used, THEN aligning to different S/Gs may be desired to limit risk of water hammer in depressurized feed lines.
* IF another feedwater source is established to the same S/G(s), THEN control steam dump to allow S/G pressure to increase and recover S/G level, while avoiding an uncontrolled cooldown.
L9 Return to FSG-3 UNIT 1, ALTERNATE LOW PRESSURE FEEDWATER Step 7.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 90 of 96 I STEP, , ACTION/EXPECTED RESPONSE I --------------------'--------------------------1 RESPONSE NOT OBTAINED ATTACHMENT M (Page 1 of 7) ASYMMETRICAL COOLDOWN NOTES
* Boration will take approximately 3 hours to complete when utilizing the BAST, approximately 6 hours for the RWST.
* Pressurizer level should be allowed to rise towards the end of the boration to support the cooldown Ml Initiate RCS Boration
* OP-SB, Blender Operation/
Dilution/
Boration
* FSG-8 Unit 1, Alternate RCS Boration M2 Check CST Level: a. CST Level -GREATER THAN 15.75 ft. b. CST Level -GREATER THAN 4 ft. a. IF an ELAP is in progress AND CST is Available, THEN perform FSG-6, ALTERNATE CST MAKEUP while continuing with this procedure.
: b. IF an ELAP NOT in progress, THEN switch to alternate AFW suction supply per AOP-23, UNIT 1, ESTABLISHING ALTERNATE AFW SUCTION SUPPLY while continuing with this procedure.
IF an ELAP in progress, THEN perform FSG-2, ALTERNATE AFW SUCTION SOURCE while continuing with this procedure.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 91 of 96 I STEP I l ___ A_c_T_I_o_N_f_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
___ ..,I ______ RE_s_P_o_N_s_E_N-:.._o
__ T_o_B_T_A_I_NE_o
_____ _ ATTACHMENT M (Page 2 of 7) ASYMMETRICAL COOLDOWN **********************************************************
M3 Check Intact S/G Level: a. S/G level -GREATER THAN [51%] 32% b. Control AFW Flow To Maintain S/G Levels Between -[51%] 32% and 63% a. Maintain maximum AFW flow until level is greater than [51%] 32% in at least one S/G 1) IF an ELAP is NOT in progress, THEN perform ATTACHMENT K, ESTABLISHING HEAT SINK. 2) IF an ELAP is in progress AND CST is NOT available, THEN perform FSG-2, ALTERNATE AFW SUCTION SOURCE. 3) IF an ELAP is in progress and an alternate low pressure feedwater source is required due to lP-29 unavailability, THEN perform the following:
a) IF 2P-29 AND associated suction source is available, THEN perform FSG-15 UNIT 1, CROSS TIE AFW. b) IF 2P-29 AND associated suction source NOT available, THEN establish low pressure feedwater per ATTACHMENT L, ESTABLISHING LOW PRESSURE FEEDWATER FLOW. **********************************************************
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 92 of 96 I STEP I I ____ A_c_T_I_o_N_f_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
___ ...,I ______ RE_s_P_o_N_s_E_N-:.._o
__ T_o_B_T_A_I_NE_D
_____ ----1 ATTACHMENT M (Page 3 of 7) ASYMMETRICAL COOLDOWN M4 Check DC Bus Loads: a. Check vital instrumentation
-AVAILABLE
* Any DC Bus voltage greater than 107.5 Vdc
* Vital instruments
-REQUIRED INSTRUMENTS AVAILABLE:
o Red Instrument Bus Powered o White Instrument Bus Powered a. Perform FSG-7 UNIT 1, LOSS OF VITAL INSTRUMENTATION OR CONTROL POWER while continuing with this procedure.
**********************************************************
M5 Maintain RCS Inventory:
: a. Check PZR level -GREATER THAN [32%] 13% a. Maintain level per FSG-1 Unit 1, LONG TERM RCS INVENTORY CONTROL. **********************************************************
M6 Check Boration Complete:
: a. Check RCS borated to target value. b. Time completed M7 Check Elapsed Time Since Boric Acid Addition Was Completed:
: a. Check one hour has elapsed since M6 .b a. DO NOT CONTINUE until one hour has elapsed.
POINT BEACH NUCLEAR PLANT E:MERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 93 of 96 l STEP! I.__ ___ A_c_T_I_o_N_f_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
_ ___,_-,1I
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__ N __ o __ T_o_B_T_A_INE
__ o _____ --1 ATTACHMENT M (Page 4 of 7) ASYMMETRICAL COOLDOWN NOTES
* Depressurization of S/G will result in SI actuation.
SI should be reset to permit manual loading of equipment on AC safeguards bus.
* PZR level may be lost and reactor vessel upper head voiding may occur due to depressurization of S/G. Depressurization should not be stopped to prevent these occurrences.
MS Depressurize Intact S/G to 320 psig: a. Check S/G narrow range levels greater than [51%] 32% in available S/G a. Perform the following:
: 1) Maintain maximum AFW flow until narrow range level greater than [51%] 32% in available S/G 2) IF an ELAP is in progress and an alternate low pressure feedwater source is required due to lP-29 unavailability, THEN perform the following:
a) IF 2P-29 AND associated suction source is available, THEN perform FSG-15 UNIT 1, CROSS TIE AFW. b) IF 2P-29 AND associated suction source NOT available, THEN establish low pressure feedwater per ATTACHMENT L, ESTABLISHING LOW PRESSURE FEEDWATER FLOW. 3) WHEN narrow range level is greater than [51%] 32% in at least one S/G, THEN continue with Step M8.b. (Step MS, continued on next page)
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 94 of 96 I STEP I I.__ ___ A_c_T_I_o_N_IE_x_P_E_c_T_E_n_RE_s_P_o_N_s_E
___ .._I ______ RE_s_P_o_N_sE_=N=O=T-o_B_T_A_I_NE
__ n _____ ---1 ATTACHMENT M (Page 5 of 7) ASYMMETRICAL COOLDOWN (Step MS. continued from previous page) b. Manually dump steam using S/G ADVs to maintain cooldown rate in RCS cold legs less than 25&deg;F/hr. o lMS-2016 o lMS-2015 c. Check S/G pressures
-LESS THAN 320 psig d. Manually control S/G ADVs at 320 psig: o lMS-2016 o lMS-2015 M9 Check SI Signal Status:
* SI -HAS BEED ACTUATED MlO Reset SI Mll Isolate SI Accumulators
: a. Perform FSG-10 UNIT 1, PASSIVE RCS INJECTION ISOLATION.
: b. Verify Accumulators
-ISOLATED
* 1SI-841A
* 1SI-841B b. Locally dump steam o lMS-2016 o lMS-2015 c. Perform the following:
: 1) S/G pressures decrease to less than 320 psig, THEN go to Step M8.d. d. Locally control S/G ADVs to maintain S/G pressure at 320 psig IF S/G pressure can NOT be maintained at 320 psig, THEN perform FSG-9 UNIT 1, LOW DECAY HEAT TEMPERATURE CONTROL to stop the uncontrolled cooldown POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 95 of 96 I STEP! I.__ ___ A_c_T_I_o_N_/_E_x_P_E_c_T_E_n_RE_s_P_o_N_s_E
___ .... I ______ RE_s_P_o_N_s_E_N-:._o
__ T_o_B_T_A_INE_n
_____ ----1 ATTACHMENT M (Page 6 of 7) ASYMMETRICAL COOLDOWN NOTE PZR level may be lost and reactor vessel upper head voiding may occur due to depressurization of S/G. Depressurization should not be stopped to prevent these occurrences.
Ml2 Depressurize Intact S/G to 150 psig: a. Check S/G narrow range levels greater than [51%] 32% in available S/G b. Manually dump steam using S/G ADVs to maintain cooldown rate in RCS cold legs less than 25&deg;F/hr. o lMS-2016 o lMS-2015 a. Perform the following:
: 1) Maintain maximum AFW flow until narrow range level greater than [51%) 32% in available S/G 2) IF an ELAP is in progress and an alternate low pressure feedwater source is required due to lP-29 unavailability, THEN perform the following:
a) IF 2P-29 AND associated suction source is available, THEN perform FSG-15 UNIT 1, CROSS TIE AFW. b) IF 2P-29 AND associated suction source NOT available, THEN establish low pressure feedwater per ATTACHMENT L, ESTABLISHING LOW PRESSURE FEEDWATER FLOW. b. Locally dump steam (Step M12 continued on next page)
L_ POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 69 Page 96 of 96 I STEP I I.__ __ A_c_T_I_o_N_I_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
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___ o __ T_o_B_T_A_I_NE_o
_____ ---1 ATTACHMENT M (Page 7 of 7) ASYMMETRICAL COOLDOWN (Step Ml2. continued from previous page) c. Check S/G pressures
-LESS THAN 150 psig d. Manually control S/G ADVs at 150 psig: o lMS-2016 o lMS-2015 M13 Go To Step 70. -END-c. Perform the following:
: 1) WHEN S/G pressures decrease to less than 150 psig, THEN go to Step Ml2.d. d. Locally control S/G ADVs to maintain S/G pressure at 150 psig. IF S/G pressure can NOT be maintained at 150 psig, THEN perform FSG-9 UNIT 1, LOW DECAY HEAT TEMPERATURE CONTROL to stop the uncontrolled cooldown.
FOLDOUT PAGE FOR ECA-0.0 UNIT 1 1 POWER RESTORATION CRITERIA o IF power is restored to any 480 Vac safeguards bus prior to placing ECCS components in pull out, THEN go to Step 46. o IF power is restored to any 480 Vac safeguards bus after ECCS components have been placed in pull out, THEN go to Step 70. 2 AFW SUPPLY SWITCHOVER CRITERIA IF CST level lowers to less than 4 ft., THEN switch to alternate AFW suction supply per AOP-23 UNIT 1, ESTABLISHING ALTERNATE AFW SUCTION SUPPLY. 3 FAULTED S/G ISOLATION CRITERIA IF any S/G pressure is trending lower in an uncontrolled manner OR any S/G completely depressurized, THEN the following may be performed:
: a. Isolate feed flow to faulted S/G. b. Maintain total feed flow greater than or equal to 230 gpm until narrow range level in at least one S/G is greater than [51%] 32% 4 ADVERSE CONTAINMENT CONDITIONS
,!!: any condition listed below occurs, THEN Environmentally Qualified (EQ) equipment and adverse containment setpoint values in brackets [ ], shall be used: o Containment pressure -GREATER THAN 5 psig o Containment radiation level -GREATER THAN OR EQUAL TO l.OE+4 R/HR o Integrated dose to containment
-GREATER THAN OR EQUAL TO 3.5E+4 R APPLICABLE ONLY DURING ELAP CONDITIONS 5 ALTERNATE AFW SUCTION CST is NOT Available OR CST level -LESS THAN 4 feet, THEN perform FSG-2, ALTERNATE AFW SUCTION SOURCE 6 ALTERNARE LOW PRESSURE FEEDWATER FLOW Perform ATTACHMENT L if TDAFW flow is lost and is NOT immediately recoverable after Step 5 is performed.
7 LOSS OF VITAL INSTRUMENTATION OR CONTROL POWER Perform FSG-7, LOSS OF VITAL INSTRUMENTATION OR CONTROL POWER if ELAP is in progress and EITHER condition listed below occurs: o DC Bus Voltages are less than 111 Vdc o Required vital instruments can NOT be energized 8 ALTERNATE CST MAKEUP Perform FSG-6, ALTERNATE CST MAKEUP if CST level -LESS THAN 15.75 feet and ALL conditions listed below occur: CST is available AFW Flow has been verified POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION A. PURPOSE AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 1 of 18 1. This procedure provides directions to respond to a diesel malfunction, when an automatic start signal is present and the diesel does not respond as required.
: 2. This procedure is applicable for all plant conditions.
B. SYMPTOMS OR ENTRY CONDITIONS
: 1. Safeguard bus lA-05 has experienced an undervoltage and the aligned diesel has failed to energize the bus as required.
C. REFERENCES
: 1. EC 283586, Transition to 10 CFR 50.48(c) -NFPA 805 From App R 2. EPM Report R2168-1003C-001 POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 2 of 18 jsTEPI I ACTION/EXPECTED RESPONSE I ,__ ___________________
..._ ______________________
___. RESPONSE NOT OBTAINED NOTE When closing electrical breakers remotely, the control switches should be held in the closed position for approximately 2 seconds to ensure UV relays have time to energize.
1 2 Check Plant Stable Per AOP-18B Unit 1, TRAIN "B" EQUIPMENT OPERATION Check Unit 1 -AT POWER IF NOT previously performed, THEN go to AOP-18B Unit 1, TRAIN "B" EQUIPMENT OPERATION while continuing with this procedure as a secondary priority.
IF Unit 1 shutdown AND both main feed pumps secured, THEN locally bypass the AMSAC actuation circuit using key #81: a. In Cable Spreading Room at panel 1N16, place bypass switch to -BYPASS b. Go to Step 6. NOTE When 1Y06 is deenergized, outward rod motion is blocked in both auto and manual modes. 3 Shift Control Rod Bank Selector Switch To -MANUAL 4 Shut MSR Steam Supply Valves: a. Adjust controller output to -ZERO
* lHC-2085 5 Shift IRPI To Alternate Power Supply: a. Place Rod Position Indication Power Transfer switch in -ALT 1Y02 POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 3 of 18 !sTEPI ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 6 Energize Bus lA-05 From Diesel G-01: 7 a. Check annunciator UNIT 1 4.16 kV BUS LOCKOUT -CLEAR
* CO2 D 3-4 b. Check G-01 to bus lA-05 breaker -IN AUTO
* 1A52-60 c. Check G-01 -RUNNING d. Ensure lA-03 to lA-05 bus tie breaker -OPEN
* 1A52-57 e. Check G-01 to bus lA-05 breaker -CLOSED
* 1A52-60 Check Bus lA-05 -ENERGIZED 8 Refer To TS 3.8, Electrical Power Systems 9 Go To AOP-18A Unit 1, TRAIN "A" EQUIPMENT OPERATION
: a. b. C. e. Perform the following:
: 1) At cubicle lA00-61, check 4.16 kV bus lockout on lA-05. 2) IF lA-05 is locked out, THEN: a) Consult with Maintenance to determine and correct cause. b) Return to Procedure And Step In Effect. Go to Step 10. Start G-01: 1) Ensure G-01 diesel mode selector switch -IN AUTO. 2) Turn G-01 diesel generator control switch to -START. 3) IF G-01 will NOT start, THEN go to Step 10. Perform the following:
: 1) Try to auto-close breaker by placing control switch to trip position and then release. 2) IF breaker will NOT auto-close, THEN: a) Place G-01 mode selector switch to -EXERCISE.
b) Turn synch switch for G-01 to bus lA-05 breaker -ON. c) Manually close G-01 to bus lA-05 breaker.
* 1A52-60 Go to Step 10.
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION I STEP' I ACTION/EXPECTED RESPONSE I CAUTION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 4 of 18 RESPONSE NOT OBTAINED In fire scenarios, G-02 is susceptible to automatic tripping on overload when it is supplying both lA-05 and 2A-05 due to spurious equipment operations.
10 Energize Bus lA-05 From Diesel G-02: a. Check G-02 -RUNNING b. Ensure lA-03 to lA-05 bus tie breaker -OPEN
* 1A52-57 c. Ensure G-01 to bus lA-05 breaker -IN PULLOUT
* 1A52-60 d. Using key #43, unlock and place G-02 to bus lA-05 breaker control switch in AUTO
* 1A52-66 e. Check G-02 to bus lA-05 breaker -CLOSED
* 1A52-66 a. e. Start G-02: 1) Ensure G-02 diesel mode selector switch -IN AUTO. 2) Turn G-02 diesel generator control switch to -START. 3) IF G-02 will NOT start, THEN go to Step 14. Perform the following:
: 1) Try to auto-close breaker by placing control switch to trip position and then release. 2) IF breaker will NOT auto-close, THEN: a) Place G-02 mode selector switch to -EXERCISE.
b) Turn synch switch for G-02 to bus lA-05 breaker -ON c) Manually close G-02 to bus lA-05 breaker.
* 1A52-66 I POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 5 of 18 I STEP 11 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I '---------------------L--------------------------'
11 Check Bus lA-05 -ENERGIZED 12 Refer To TS 3.8, Electrical Power Systems 13 Go To AOP-18A Unit 1, TRAIN "A" EQUIPMENT OPERATION 14 Check Bus H-02 -ENERGIZED Go to Step 14. IF bus H-01 energized OR bus H-03 ~ergized, THEN perfor;;:;-the following:
: a. IF annunciator 13.8 kV BUS OR FEEDER LOCKOUT in alarm, THEN go to Step 16.
* CO2 E 2-9 b. Ensure bus H-02 normal feed -OPEN
* H52-20 c. Ensure H-03 to H-01 bus tie breaker -CLOSED
* H52-31 d. Turn synch switch for H-02 to H-01 bus tie breaker -ON e. Close H-02 to H-01 bus tie breaker.
* H52-21 IF bus H-02 can NOT be energized, THEN go to Step 16.
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 6 of 18 lsTEPI I ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I ...._ __________________
..__ _____________________
__. 15 Restore Power To lA-03: a. Check annunciator UNIT 1 4.16 kV BUS LOCKOUT -CLEAR
* CO2 D 3-4 b. Check annunciator lX-04 LOW VOLTAGE STATION AUX TRANS LOCKOUT -CLEAR
* CO2 D 1-7 c. Reset and close bus H-02 feed to lX-04 breaker
* H52-22 d. Turn synch switch for bus lA-03 normal feed breaker -ON e. Reset and close bus lA-03 normal feed breaker
* 1A52-36 a. b. c. e. Perform the following:
: 1) At cubicle 1A52-40, check 4.16 kV bus lockout on lA-03. 2) IF lA-03 is locked out, THEN select emergency diesel power supply: o ATTACHMENT A, G-01 LOCAL MANUAL START o ATTACHMENT B, G-02 LOCAL MANUAL START Go to Step 16. Go to Step 16. Go to Step 16.
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 7 of 18 I STEP 11 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I ..._ __________________
..__ _____________________
__. 16 Check Bus lA-03 -ENERGIZED Perform the following:
: a. IF bus lA-01 is powered from lX-02, THEN energize bus lA-03 from lA-01: 1) Ensure bus lA-03 normal feed -OPEN
* 1A52-36 2) Ensure lA-03 to 2A-03 bus tie breaker -OPEN
* 1A52-40 3) Turn synch switch for lA-03 to lA-01 bus tie breaker -ON 4) Close lA-03 to lA-01 bus tie breaker.
* 1A52-37 5) Turn synch switch for lA-03 to lA-01 bus tie breaker -OFF 6) IF power is restored to lA-03, THEN go to step 17. b. IF bus 2A-03 is powered from 2X-04, THEN energize bus lA-03 from 2A-03: 1) Ensure bus lA-03 normal feed -OPEN
* 1A52-36 2) Ensure lA-03 to lA-01 bus tie breaker -OPEN
* 1A52-37 3) Turn synch switch for lA-03 to 2A-03 bus tie breaker -ON 4) Close lA-03 to 2A-03 bus tie breaker.
* 1A52-40 5) Turn synch switch for lA-03 to 2A-03 bus tie breaker -OFF c. IF lA-03 can NOT be powered from lA-01 or 2A-0~THEN select emergency diesel power supply: o ATTACHMENT A, G-01 LOCAL MANUAL START o ATTACHMENT B, G-02 LOCAL MANUAL START POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 8 of 18 I STEP 11 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I '----------------------L-------------------------1 17 Energize lA-05 From lA-03: 18 a. Ensure G-01 to bus lA-05 breaker -OPEN
* 1A52-60 b. Ensure G-02 to bus lA-05 breaker -OPEN
* 1A52-66 c. Turn synch switch for lA-03 to lA-05 bus tie breaker -ON d. Reset and close lA-03 to lA-05 bus tie breaker
* 1A52-57 Check Bus lA-05 -ENERGIZED 19 Refer To TS 3.8, Electrical Power Systems 20 Go To AOP-18A Unit 1, TRAIN "A" EQUIPMENT OPERATION
: d. Select emergency diesel power supply: o ATTACHMENT A, G-01 LOCAL MANUAL START OR o ATTACHMENT B, G-02 LOCAL MANUAL START Return to Step 6. -END-POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 9 of 18 'STEP' I ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I ____________________
...._ ______________________
__, Al A2 Check Green "Power On" Light -ENERGIZED
* Panel C-64A
* Panel C-34 Check Overspeed Trip Alarms -CLEAR
* Panel C-64A
* Panel C-34 ATTACHMENT A (Page 1 of 5) G-01 LOCAL MANUAL START IF green light NOT lit, THEN transfer control power to alternate source: a. At PAB 8' elevation South of Unit 2 charging pumps, direct PAB operator to shut switch D31-01. b. At C-78, shift to alternate power by swapping paired breakers:
* For annunciators, open breaker 1 and close breaker 2.
* For start circuit 1, open breaker 3 and close breaker 4.
* For control power, open breaker 5 and close breaker 6.
* For field flash, open breaker 7 and close breaker 8. Reset mechanical overspeed trip and alarms: a. Place mode selector switch in -LOCAL b. Reset mechanical overspeed trip. c. Place mode selector switch in -AUTO POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 10 of 18 I STEP 11 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I ,__ ___________________
..__ ______________________
__, ATTACHMENT A (Page 2 of 5) G-01 LOCAL MANUAL START A3 Emergency Start G-01: A4 a. At C-34A, place local/remote transfer switches to -LOCAL
* Transfer switch No. 1
* Transfer switch No. 2 b. At C-34A, start G-01 by depressing EMERGENCY START push-button At C-64, Check G-01 Speed -GREATER THAN OR EQUAL TO 900 RPM b. IF G-01 will NOT emergency start, THEN manually start G-01: 1) At C-64, place mode selector switch in -LOCAL START. 2) At C-64, depress and hold ENGINE START push-button until engine speed rises to idle. 3) At C-64, raise engine speed to 900 rpm by depressing idle release push-button.
: 4) IF G-01 can NOT be started, THEN do not continue and inform Control Room of G-01 status. Perform the following:
: a. At C-64, place governor mode switch to -"HYO" b. IF diesel speed NOT greater than or equal to 900 rpm, THEN at C-64, raise speed using hydraulic governor control switch.
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 11 of 18 I STEP 11 ACTION/EXPECTED RESPONSE I .._-------------------~------------------------1 RESPONSE NOT OBTAINED A5 ATTACHMENT A (Page 3 of 5) G-01 LOCAL MANUAL START CAUTION Rubber gloves with leather protectors and personnel safety equipment are required to locally operate FFC relay. Check G-01 Frequency
-BETWEEN 59.7 Hz HZ AND 60.3 Hz Perform the following:
: a. IF field NOT flashed, THEN in C-34, manually actuate relay FFC for 3 seconds by pushing yellow tab on relay actuator located behind terminal connections.
: b. At C-64, ensure 43/G-01-GOV, GOl EOG Mode Selector Switch in HYO, THEN adjust frequency using the hydraulic governor control switch. c. IF hydraulic governor control switch NOT functional, THEN adjust frequency using SPEED control knob on faceplate of Woodward governor.
: d. IF frequency can NOT be maintained,~
locally shutdown G-01: 1) At C-64, push both engine stop pushbuttons.
: 2) Pull fuel supply cut-off valve operator.
: 3) At C-34A, place output breaker in pullout:
* 1A52-60 4) Go to ATTACHMENT B, G-02 LOCAL MANUAL START.
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 12 of 18 I STEP 11 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I ...__ ________________
-'---------------------...J A6 ATTACHMENT A (Page 4 of 5) G-01 LOCAL MANUAL START Contact Control Room To Check G-01 Voltage -BETWEEN 4050 VAC AND 4300 VAC IF field is NOT flashed, THEN in C-34, manually actuate relay FFC for 3 seconds by pushing yellow tab on relay actuator located behind terminal connections.
Contact Control Room to maintain voltage between 4050 Vac and 4300 Vac by adjusting diesel loading. a. voltage can NOT be maintained, THEN locally shutdown G-01: b. Push both engine stop push-buttons.
: c. Pull fuel supply cut-off valve operator.
: d. At C-34A, place output breaker in pullout.
* 1A52-60 e. Go to ATTACHMENT B, G-02 LOCAL MANUAL START.
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 13 of 18 I STEP 11 I ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I ...._ ___________________
L..-------------------------1 ATTACHMENT A (Page 5 of 5) G-01 LOCAL MANUAL START A7 Energize Bus lA-05 From Normal Diesel Supply G-01: a. Check G-01 -RUNNING b. Locally ensure lA-03 to lA-05 bus tie breaker -OPEN
* 1A52-57 c. Locally ensure G-01 to 2A-05 bus tie breaker control switch -IN PULLOUT
* 1A52-73 d. Locally ensure G-01 to bus lA-05 breaker control switch -IN AUTO
* 1A52-60 e. Locally check G-01 to bus lA-05 breaker -CLOSED
* 1A52-60 AB Go To AOP-18A Unit 1, TRAIN "A" EQUIPMENT OPERATION
: a. b. c. d. e. -END-Go to ATTACHMENT B, G-02 LOCAL MANUAL START. Go to ATTACHMENT B, G-02 LOCAL MANUAL START. Go to ATTACHMENT B, G-02 LOCAL MANUAL START. Place G-01 to bus lA-05 breaker control switch -IN AUTO Locally: 1) Try to auto-close breaker by placing control switch to trip position then release. 2) IF breaker will NOT auto-close, THEN manually close breaker control switch.
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 14 of 18 I STEP 11 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I '-------------------'----------------------'
ATTACHMENT B (Page 1 of 5) G-02 LOCAL MANUAL START Bl Dispatch Operator With Key #43 To G-02 B2 B3 Check Green "Power On" Light -ENERGIZED
* Panel C-65A
* Panel C-35 Check Overspeed Trip Alarms -CLEAR
* Panel C-65A
* Panel C-35 IF green light NOT lit, THEN transfer control power to alternate source: a. At PAB 8' elevation South of Unit 2 charging pumps, direct PAB operator to shut switch D41-01. b. At C-79, shift to alternate power by swapping paired breakers:
* For annunciators, open breaker 1 and close breaker 2.
* For start circuit 1, open breaker 3 and close breaker 4.
* For control power, open breaker 5 and close breaker 6.
* For field flash, open breaker 7 and close breaker 8. Reset mechanical overspeed trip and alarms: a. Place mode selector switch in -LOCAL b. Reset mechanical overspeed trip. c. Place mode selector switch -IN AUTO POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 15 of 18 I STEP 11 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I ____________________
....._ ______________________
___, ATTACHMENT B (Page 2 of 5) G-02 LOCAL MANUAL START B4 Emergency Start G-02: BS a. At C-35A, place local/remote transfer switches to -LOCAL
* Transfer switch No. 1
* Transfer switch No. 2 b. At C-35A, start G-02 by depressing EMERGENCY START push-button At C-65, Check G-02 Speed -GREATER THAN OR EQUAL TO 900 RPM b. IF G-02 will NOT emergency start, THEN manually start G-02: 1) At C-65, place mode selector switch in -LOCAL START. 2) At C-65, depress and hold ENGINE START push-button until engine speed rises to idle. 3) At C-65, raise engine speed to 900 rpm by depressing idle release push-button.
: 4) IF G-02 can NOT be started, THEN do not continue and inform Control Room of G-02 status. Perform the following:
: a. At C-65, place governor mode switch to -"HYO" b. IF diesel speed~ greater than or equal to 900 rpm, THEN at C-65, raise speed using hydraulic governor control switch.
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 16 of 18 lsTEPI I ACTION/EXPECTED RESPONSE I ...._ ___________________ ,L_ ______________________
__. RESPONSE NOT OBTAINED B6 ATTACHMENT B (Page 3 of 5) G-02 LOCAL MANUAL START CAUTION Rubber gloves with leather protectors and personnel safety equipment are required to locally operate FFC relay. Have the control room check G-02 Frequency
-BETWEEN 59.7 Hz AND 60.3 Hz Perform the following:
: a. IF field NOT flashed, THEN in C-35, manually actuate relay FFC for 3 seconds by pushing yellow tab on relay actuator located behind terminal connections.
: b. At C-65, ensure 43/G-02-GOV, G02 EOG Mode Selector Switch in HYO, THEN adjust frequency using the hydraulic governor control switch. c. IF hydraulic governor control switch NOT functional, THEN adjust frequency using SPEED control knob on faceplate of Woodward governor.
: d. IF frequency can NOT be maintained, THEN locally shutdown G-02: 1) Push both engine stop pushbuttons.
: 2) Pull fuel supply cut-off valve operator.
: 3) At C-35A, place output breaker in pullout:
* 1A52-66 4) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 17 of 18 'STEP' ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED B7 ATTACHMENT B (Page 4 of 5) G-02 LOCAL MANUAL START Check G-02 Voltage -BETWEEN 4050 VAC AND 4300 VAC ,!!'. field is NOT flashed, THEN in C-35, manually actuate relay FFC for 3 seconds by pushing yellow tab on relay actuator located behind terminal connections.
Contact Control Room to maintain voltage between 4050 Vac and 4300 Vac by adjusting diesel loading. IF voltage can NOT be maintained, THEN locally shutdown G-02: a. Push both engine stop push-buttons.
: b. Pull fuel supply cut-off valve operator.
: c. At C-35A, place output breaker in pullout.
* 1A52-66 d. Return to Procedure And Step In Effect.
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE TRAIN "A" SAFEGUARDS BUS RESTORATION AOP-19A Unit 1 SAFETY RELATED Revision 13 Page 18 of 18 I...._ ___ A_c_T_I_o_N_I_E_x_P_E_c_T_E_D_RE_s_P_o_N_s_E
___ ...,I ______ RE_s_P_o_N_s_E_N_o_T_o_B_T_A_I_NE_D
_____ _.I ATTACHMENT B (Page 5 of 5) G-02 LOCAL MANUAL START BB Energize Bus lA-05 From Alternate Diesel Supply G-02: a. Check G-02 -RUNNING b. Locally ensure lA-03 to lA-05 bus tie breaker -OPEN
* 1A52-57 c. Locally ensure G-02 to 2A-05 bus tie breaker control switch-IN PULLOUT
* 2A52-67 d. Locally unlock and place G-02 to bus lA-05 breaker control switch -IN AUTO
* 1A52-66 e. Locally check G-02 to bus lA-05 breaker -CLOSED
* 1A52-66 B9 Go To AOP-lBA Unit 1, TRAIN "A" EQUIPMENT OPERATION
: a. b. C. e. -END-Return to Procedure And Step In Effect. Return to Procedure And Step In Effect. Return to Procedure And Step In Effect. Locally: 1) Try to auto-close breaker by placing control switch to trip position then release. 2) IF breaker will NOT auto-close, THEN manually close breaker control switch.
ATTACHMENT 5 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 286, ADOPTION OF EMERGENCY ACTION LEVEL SCHEME PURSUANT TO NEI 99-01 REVISION 6, "DEVELOPMENT OF EMERGENCY ACTION LEVELS FOR NON-PASSIVE REACTORS" UPDATED PBNP EAL SCHEME WALLBOARDS 3 pages follow R Abnormal R-, Level* I R-, Effluent 1 R-, Efflue nt 2 ........ olO,,MOUI rNloacltvlty ,HUlllng 6n offstl* doH ..... tt,-n 1,000 mrem TEDE or 1 , 000 ,,...,. ~CHd C OE r on 36 i 1 2 J 4 5 6 I DEF RG 1.1 Reading on A N Y Table R-1 effluent radiation monitor greater than column "GE" f or 15 minutes or longe r. R G 1.2 Dose asse s sment using a ctual meteo r ology indica t es doses greater than 1000 m rem TEDE o r 5000 mrem thyroid COE at or beyond the SITE BOUNDARY R G 1.3 Field survey results Indicate EITH ER of the f ollowing a t or beyond the SITE BOUNDARY:
Closed window dose rates greater than 1000 mR/hr expected to continue for greater th a n or equal to 60min.
* Analyses of field survey samples indicate thyro i d COE gr e ater than 5000 mrem fo r 60 min. of inhalaUon. "-NH o f gaseo u , r.cfloadlv lty ,nutting I n offsh doH greater than100mr.mTEDEo r IIOO-emth.,...,ldCDE
'""-33 1 2 l 3 I 4 I 5 I 6 I DEF RS 1.1 Reading on AN Y Table R-1 effluent radiation mo n ito r greater than column "SAE" for 15 minutes or longer. RS1.2 Dose assessment using actual meteorology indicates doses greater than 100 m rem TEDE or 500 mre m thyroid COE a t o r beyond the SITE BOUNDARY R S 1.3 Field survey results indicate E I TH ER of the following at or beyond the SITE BOU N DARY:
* Closed window dose rates greater than 1 00 mR/hr expected to conUnue for g reater than o r equal to 60 min.
* Analyses o f fiel d surv e y samples indicate th y roi d COE greater than 500 mrem for 60 min. of inhalation Sl*!lfuetpool lewlc arw,o 1 1N rHtoredt o at IH,tthelop Spanl fueJ~leval atthetopoftheP\>>I r aclt,[pg. SS] o lthefuel r ack, for IO ml11U1e1 or longe r{pg. S8) 516 IDEF 1 12 J3 J4 [5 1 6 IDEF R G2.1 R S2.1 Spent fuel pool level cannot be restored to at least 40 ft. 8 in. Lowering of spent fuel pool level to 40 fl 8 in. fo r 60 rrinutes or lo n ger. Table R-1 Efflue n t Monttor ClaHlfteatlon Threshold s I rradiated I I,--,.------------.---~~-----,-----.-----
.. Fuel E vent 3 Area Radiatio n Levels Monitor 1(2}-RE-307 C TM NT Pu i g, 8r h ,u$t M id Range 0.1 1(2)-R E-309 C TMN T Pu r ge Elthaust High R*nge Gu _flff'/_ ..........
., __ tion 2*RE-305 CTMN T Pur g e 8rhau1t Low Ra~* Ga1 --~g,o-GSbuMi'lgv...iutionln-ellan 2-RE-307 CTMNT Purge Exh a ust M id Range 0.1 -bod'i-*""GS~V$"0tilllionlfl-llion 2*RE*307 CTMNT Purge Exhaust M id Range G11 -onftGSbw:ling&#xa5;W'llilltiorllnop,tf1Ciool 2-R E-307 CTMNT Pu1g-e Ex h a ust M id R ange Gu -onh<klnlad_d__,,,,.. ~;;,.~T=~~:-..::.~=*Gn !!~~=~~:.i::lghR,ngeGH 2*R E*:QI CT M NT Purge Exhaust High R*ng* Gu -onft l cwotd_ol....._llvnont R E-315 AB Exh a ull Low Ra nge Gu G E 6.0E+2pCllcc 4.0E+2 pCllcc 1.2E+3&#xb5;Cllcc 1.0E+2pCllcc 1.0E+211Cllcc SAE A l e rt UE 6.0E+1j.l(:llcc 6.0E+O~cc 6.0E+1 pCllcc 6.0E+opCUcc 9.o1E-3j.l(:llcc
-1.0E+1 j.l(:llcc
-1.0E+Oj.l(:llcc 9.4E-Jj.l(:llcc 2.8E*2&#xb5;Cllcc t.OE+t &#xb5;Cllcc 4.0E+t ..,CUcc -1.0E+Oj.l(:llcc 1.2E+2&#xb5;Cllcc 1.2 E+1 &#xb5;Cllcc 1.0E+1 .,Cllcc 5.4E*3pCl/cc 1.0E+t pCllcc 1.0E+OpCllcc 5.4E*311Cllcc 1.0 E+111Clloc 1.0E+011Cilcc 8.4E-311Cllcc 1.6E+111Clloc 1.6 E+011Cllcc e .. o1E.J11Cllcc
!---1-----tt i 1(2}-RE*229 Servtc. Waler OY1rboard
::; E 1 C o nnoem e nt IS FSI Bound*ry H H aza r ds 1 Secu r i ty 2 S e ism ic Event 3 Natu r a l or Tec h. Haza rd 4 Fire 5 Hazardous Ga ses 6 Con tr o l Room Ev ac u ation 7 ED Judgment HOS TI LE A C TI O N ru ul llng In 1 0 11 o f ph y1k:a1 co ntr ol o l t he fac lll ty (pg.HJ HOSTILE A C TI O N within !he PR OTE C TED A REA [pg. HJ DEF 1 I 2 I J I 4 I 5 I 6 I DEF H G 1.1 A HOSTILE ACTJON is occurri ng o r has occu r re d within t he PRO T ECTED AREA as repo rt ed by the Security Shift Supervisor ANO EITH ER of the following has occurred: ANY of the follO'Ning s a fety functions cannot be controlled or maintain ed Reactivity control Core cooling RCS heat removal OR
* Damage to spent fuel has occurred or is IMMINENT Ot h er co nd i t lo~ ex l, t i ng t hat In th e J uclgmei,t o l t he Eme rg e i,c y Co o rd i nato r w ar ra n t dect.ra tlo i, o l Ge11eral E mer genc y [pg. 101] DEF H G7.1 Ot h er cond it ions exist which i n t he judgment of the Emergency Director indicate th a t e ven ts a re In p ro g ress or have occurred which involve actual or I MM INENT substanUal cor e degrada t ion or mel t ing w ith potential for los s of con ta inment integ r ity o r HOSTILE ACTION t ha t results i n an actual loss of physical control of the f a cility. Rele a ses c a n b e reasonably expected to exceed EPA Pro t ective Action Guideline exposu re l ev els o ffs ite for more than the immediate site ar e a. A HOSTILE ACTION is occurring o r has occurred within the PROTECTED AREA a s report ed b y the Security Shift Supervisor l 11ab llf ty to co ntrol a key H lety fu nc ti o n from out1 l d e th e Co n l 1 ol R oo m (pg. 97] 1 2 3 D E F HS 6.1 An event h a s resulte d I n plant co n trol being transferred from the Control Room to AOP l ocal control s t ations. AN O Con t rol o f ANY of t he following k ey s a fe t y f unctions is not reestablished within 15 minutes
* Re a ctivity
* Core cooling
* RCS h e at r emoval Other cond1U o n1 e xl$l l ng that In l he judgment of the Emer g enc y Coo rdi nator w.11rrant d ec laration of Site Area E me r genc y [pg. 98) 1 I 2 I 3 I 4 I 5 I 6 I DEF HS7.1 Other conditions exist 'Mlich i n the ju d gment of the Emerge n cy Director Indicate t hat e vents a re in progr e ss or have occurre d which involve actual or lik e ly major failures of p l a nt fu nctio ns n eede d for pro t ection of t he pu b lic or HOSTILE A CTION th a t result s ln in t en ti o na l d a ma g e or malicious a cts, (1) toward site personnel or equipment th a t co u ld lea d to the li k ely fail u re o f o r , (2) t h a t preven t e ff ective access to equipment needed for the p rot e ction of the p ublic. Any r e leases are not expected t o result i n exposure levels which e xceed EPA Pro t ective Actio n Gui d eli n e exposure levels beyon d the s ite boundary. R .... N of g,i.-ou, o r ll quld rad ioactivity r Hu lllng In o ff sk* doMIINlilt..-than10 mrffl!TI:DE orl50mremth ,r ok1C DE 1nn. 27'1 1 l 2 l 3l4ls lslo e F RA1.1 Reading on ANY Table R-1 effluent radiatio n roonitor greater than column "ALERT" for 15 minutes or longer. RA1.2 Dose assessmen t using a ctual meteorology indica t es doses greater than 10 rrvem TEDE or 50 mrem thyroid COE al or beyond S I TE BOUNDARY. RA1.3 Analysis of a liquid effluent sample indicates a concentr a tion or release rate that would r esult i n doses greater than 10 mrem TEDE or 50 mrem t hyroid COE at or beyond the SITE BOUNDARY for one hou r of exposure. RA1.4 Field survey resul t s indicate EITHER of the following at or beyond the SITE BOUNDARY:
* Closed window dose r a t es greater than 10 mR/hr expected to conti n ue fo r 60 minutes or longer.
* Analyses of fi e l d s urv e y s a m p les indicat e th y roi d CO E greater than 50 rTYem for one hour of inhalation. S9"ine.nt low.Ol"g o l -':<< l eve l above, or clafmlge t o. lrradlndfuet (pg.29) 1 I 2 I 3 I 4 I 5 DEF RA2.1 Uncovery o f irradiated f uel in the REFUELING PATHWAY RA2.2 Damage to irradia t ed fuel resul t in g in a release of r a d ioactivi~
from the fuel as Indicated by a reading on ANY o f the fo!IO'Ning radiation monitors greater than the value shown:
* RE-105 SF P Area Lo w R a nge Radiation M onitor 4 R/hr
* 1(2)-RE-126 Con t ainment High Radiation Monitor 7 R/hr
* 1(2)-RE-127 Cont a inment High Radiation Monitor 7 R/hr
* 1(2rRE-128 Con ta i n ment High Radiation Monito r 7 R/hr RA2.3 L owering of s p ent fuel p ool level to 49 ft.0 in. R..:fiatlo i, levels that I MPEDE accen to ~ipnwnt: n~e11ary for normal plarrt op91atlo111.
coo l down or 1hut down{pg. 3 1 J 1 2 3 4 5 6 DEF RAJ.1 Dose rate gre a ter than 1 5 mR/hr i n ANY of the f o ll owing are a s:
* Control Room (RE-101)
* Cen t ral A lar m S t atio n AND Second a ry Alar m S1a t lon {by1u1Vey)
RA3.2 An UNPLANNED event results in radiation levels that p rohi b it or IMPEDE access t o any Table R-2 plant rooms T a bl e E*1 Cuk On-Contact D ose Rates 32 PT D S C
* VSC-24 Rel*H* o f g,-u, or l lquld rad loaietlvNy g r eat., than 2 tl 1M 1 the OOCM H rnlt , for IO mlnutH or lonaer {pg. 231 1l2 I J 1 4 l 5 l6 IDEF RU 1.1 Reading on ANY Table R-1 effluent radiation monitor greater t ha n column "UE" for 60 minutes o r longer. R U 1.2 Reading on ANY effluent radiation monitor greater than 2 times the alarm se t poi n t established by a current radioactivity discharge p ermit for 60 minutes or longe r. R U1.3 Sample an a lyses for a gaseous or liquid release indicates a concentration or rele a se rate greater than 2 times the ODCM limits for 60 minutes or longer. Unpunn.d 1 011 o l lN!ar lev e l abow lrradlalltd ru.1 [pg. 251 4 I 5 RU2.1 a. UNPLANNED water l evel drop In the REFUELING PATHWAY as indicated by ANY o f the followin g:
* Spent fuel pool low water level alarm
* Visual observation A N O b. UNPLANNED rise i n area radiation levels as i n dicated by ANY of the fol10'Ning radiation monitors.
* RE-105 SFP Area Low Range Radiation Monitor
* RE-135 SF P Ar e a High Range Radiation Monitor DEF
* 1(2) RE-102 El. 66' CONTAINMENT Low Range M onitor Tab l e R-2 Safe Op s, SID , C I D Areas Room/Area U1 VC T AIH U2VCTNH Ul PrlmaryS a mplearH U2Prtm,,rySampleare, CCWHX Room C*59a r H Pipeway2,8n.
.Ellv. Pipewa y 3,8 11.Ellv. 112832 M CCAIH Mode 3/4 15 3/4/5 J J .,, 3/4 15 ,,. '" 4 O anu,ge to a loaded Clllk CONFIN EMENT B OU NDAR Y {pg. 11) 1 2 13 1 4 1 5 1 6 IDEF EU 1.1 HSM F ront 1700 rTVern'h r l SidH 1 200 rTVern'hr Damage to a lo a ded cask CONFINEME N T BOUNDARY as End :!C:'1 Ex1 4~ :::::~ MrTl~~b :: :::'.::: Indicated by an on~ntact radiation reading greater than the ._ ___ _.._ __ .&,,;;"';;;'&deg;"';;;.".," ... ';;;;".~ .. *mlh ... , I I :~k~s shown on Table E-1 on the surface o f the spent fuel HOSTILE A CTION with i n th e OWNER C O NTROU.ED AREA o r a l r born ea tt a c kthr ea twlt hl n30 ml n vtH (pg.lt) DEF A HOSTILE ACTION is occu r ring or has occurred within the 01/vNER CO NT RO LL ED AREA as reported by the Security Shi1t Supervisor. HA1.2 A validated notification from NRC of an aircraft attack t hre a t within 30 mi n utes of t he site. T a ble H*1 Fi re A re as Control Room Containment PAB ?fik~i~t 1 ifcung Cob l e Spretiding Room X\~ ~~t\~?oo m Q.()J ,02 Roo m s EDGBuilding Service Wotl!r Pl.Imp Rooms Facade 85' Ga1e o u 1 releaH lmpedl ng a cce11 to eq uipmei,t 11eCHH r y !o r normal plant ope r a tio n*. co o l down or 1 hu t down (pg. 11) DEF H A5.1 Release of a toxic, corrosive, asphyxi a nt or fl a rrvTiable gas into any Table H-2 plant rooms or areas ANO E n try into the room or are a Is pr ohibited or Impeded. Co11trol R oom ev a c u a ti o n rHulllng I n 1 , an 1 fe r o f p l ant c ontr o l to , 1te m at e l ocatl on 1 rnn. 931 1 12 T3 T4T 5T5lD EF H A6.1 A n event has resulted in pl a nt control being t ransferred from the Control Room t o AOP l oc a l co n tro l st a tions. Other co 11d J t ion , ex is ting that In the judg ment o l t h e Emer gency C oo r dinat o r war r an t decl11at lo 11 o f ai, A l ert fpg. 94) DEF HA7.1 other conditions exist which, in the judgment of the Emergency Dir e ctor , indicate that events a r e ln progress or have occurred which Involve an act u al or p o t ential substantial degradation of the level of safety of the plant or a security event that involv e s probable li f e threatening risk to site p ersonnel o r dama g e t o site equipment b ecause o f HOSTILE ACTION. Arly releases are expected to be limited to small fractions of the EPA Protective Action Guideline expos u re l evels. Con tlrmecl SEC U RITY CO N DITIO N o r t h reat (pg. 79) HU1.1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by t he Security Shift Superviso r. H U 1.2 DEF Notification of a credible security threat directed at PBNP. H U1.3 A validated notification from the NRC providing information ofan aircraft threat. S eismi c ev e i, t g r eat er th ai, O B E level [pg. 81] D E F HU2.1 Seismic event greater t han Operating Basis Earthquake (OBE) as indicated by seismic mo n itor indication ol g round acceleration greater th an 0.06 g horizontal OR 0.04 g vertical.
H aur do u s ev en t {pg. t2] HU J.1 A tornado strike within the PROTECTED AREA HU3.2 DEF Internal room or a r ea flooding of a magnitude sufficient to require manual or automatic electrical isola t ion of a SAFETY SYSTEM component needed for the current operating mode. HU3.3 M ovement of pe r son n el within t he PROTECT E D AREA is irfl)eded due to an offsite event invoMn g hazardous materials (e.g .* an offsite chemical spill o r toxic gas release). HU 3.4 A h a zardous event t ha t results in on-site conditions sufficien t to prohibit the plant staff from a ccessing the site vi a p ersonal vehicles. HU J.5 Lake level greater than or equal to +8.0 fl (Plant elevation)
H U3.6 Pump bay level less than-15.0 ft. F I RE po t e11tlally deg rad ing t h e le vel o l ul ety o f the pla nt (pg. 8-4) HU4.1 A FIRE is n ot extinguished within 15 min. of ANY of the following FIRE detection Indications
:
* Report from the field (I.e., visual observation)
* Receipt of multiple (more than 1) fire alarms or indications
* Fiel d verification of a single fire alarm ANO DEF The FIRE is loca t ed within A N Y Table H-1 plant rooms or a reas H U4.2 Receipt of a single fire alarm with no ot h er Indications o f a FIRE ANO The FIRE is loca t ed within AfJY Table H-1 plant rooms or areas except Containment in Mo d es 1 a nd 2 ANO The existence of a FIRE is not verified within 30 minutes of al ar mreceipl HU4.3 A FIRE within th e pla nt o r lSFSI PROTECTED AREA not extinguished within 60 minutes of the initial report, al a rm or in d ication. HU4.4 A F I RE within the plan t or lS F SI PROTECTED AREA that r equires firefighting sup p ort by a n offsite fi re response agency to extinguish. T ab l e H-2 Safe Op s , SID. C I D A r eas Room/Are a U1 VCTAlea U2VCTArea U1 Pmv.ry Sample area U2 Primary Sample a rea CCWHXRoom c-59arH Piptway2,e n..e11v. Piptway3 , 8 n..Ellv. tl2B32 M CC Af aa Mode J/4 1 5 3/4/5 J ' '" 31-115 ,,. ,,. 4 -O the r condltion*
ex l s tl11g t h at In the judgment o f th e E merg e nc y C oord i nator warrant decla r a t ion of a U E [pg. HJ D EF H U7.1 Other condit i ons exist which in the judgment of the Emergency Director indicate that events are in progress or h a ve o ccurred which indicate a po t enUal degradallon of the level of s a fety of the p l a nt o r indicate a s e curity thre a t to facility protection has b een initiated. No releases of r ad io a ctive material requiring offsite response or monitoring are expected unless furthe r de gr ad a tion o f SAFETY SYSTEMS occurs. Modes: ~1~' ~' _2~11 3 .__4 __,!I.__ _s __,!I 6 1 1 DEF Point B e ach Nuclear Plant EAL Classification Matrix Page 1 of 3 ALL MODES Po\N'e r Operation S tartup Hot Standby H o t S hu tdo\oVfl Co ld Shutdown Ref u eling D ef uel ed 
,, .. , . ' .. ' .. .. . ,., .. -....... *"**--1**-..... ,...-..,""", ... ._,..,....
__ ~]IIIIIIIff GENERAL EMERGENCY SITE AREA EMERGENCY I UNUSUAL EVENT ~=:::::::::::::=:=:=:=:=:=:=:=:=
:=:= ................
1,. 1, 1 L o n o r Emo rg e n cy AC P o w er 2 lo ss o f Vital DC 1 P rolonged 10,, ot *II off*lt* a nd *It onalle A C po-l o , af egu;ardbt lH l[pg. 12f} 2 Lou o f a ll ofh[le aod a ll o n*lt* AC po-lo u l egua rd bu1e1 For 115 mi nut uorlonger (pg 12SJ ' ' S0 1.1 S S 1.1 Loss of ALL offsite a nd AL L onsite AC power t o 1(2rA05 and loss of A U offsite and All onslteAC power to 1(2)-AOS 1(2)-AD6. and 1(2)-A06 fo r 15 m inutes or longer. A N O E IT H E R:
* Restoration of at least one AC emergency bus In less than 4 hours is not likely.
* Conditions requiring entry into Core Cooling -Red Path (CSP-C. 1) are met Lon o f al l A C *ndVlt*I OC po-r ,ou r ce, for 1 5ml n utH o r I *--**-121n 1 2 S G2.1 Loss of ALL offsite alld A L L onsite AC power capability power to 1(2)-AOS and 1(2)-AOO for 15 minutes or longer. A N O Lou o f a ll Vit al D C po-r for 11 mlnutH or longer [pg. 12111) SS2.1 Ind i cated voltage is less than 115 VOC on All Vital DC busses D-01. D-02 , ~3. and D-04 for 15 minutes or longer. Lon o f all but on e A C po-r 1ourc1 lo u fe g wor d buHs for16mlnutuorlong er[pg.11'] S A 1.1 AC power capability to 1(2)-AOS AND 1(2)-A06 is reduced to a single power source for 15 minutes or longer. ANO Ally additional single power source !allure will result In a loss of ALL AC power to SAFETY SYSTEMS I Lou o l all o ff1!te AC po-r ca~bility to Hfegu.ud buHs !or 115 mi nutes o r l onge r (pg. 103] ' S U 1.1 Loss of A L L o ff site AC pov,.ier capability to 1(2}-AOS and 1 (2}-A06 for 15 minutes or longer. P o we r lndlcated voltage Is less than 115 VOC on All Vital DC bosses ~1. ~2. D-03 elld 0-04 for 15 minutes or longer. s S y s tem M a lfun ct. 3 L o ss o f C on t rol R oo m I n d ica t ions 4 R C S Act iv ity 5 R CS Le akage 6 R PS F a il ure 7 lo no f Comm. 8 CMT Fa il u re 9 H aza rd o u s Event Affecting Safety Syst.nw F FIHlo n Produ ct Bar rie r D eg radation Category A Crft lc al S afety Fu n ct io n Statu s B R C S o rS G Tube leakage C R CS Ac t i v ity I Co n t ai nme n t Rad ia t i on D Con t ainme n t In t egrity or Bypa ss E ED J udgmen t F G 1 Loss of an y two b a rriers alld Loss or Potential Loss of third barri e r (Table F-1) (pg. 63) Ta ble 5-2 Slg n l fl ca nt Tran sie nt s Automa t ic or manual runback grea t er than 25% thermal reactor power Electrical load rejection greater than 25% full electrical load Readortrip SI actuation I nability t o ,hu1 down the re actor causing a chaH1noe to co re cooling or R C S h e at removal [pg 125] SS6.1 Ar, automatic or manual tr ip d l d not shut dO\Nfl the reactor ANO All actions t o shut down the reactor have been unsuccessful AN O E ITH E R: Con d itions req u iri n g entry i nto Core Cooling -Red Pat h (CSP-C.1) are me t. Conditions requiring entry Into Heat Sink-Red Path (CSP-H.1) are met. F S 1 2 Table S.J H azardo u s Ev e nt s Seismic event (earthquake)
Internal or ex t erna l flooding event High 'Ninds or t orn a do strike FIRE EXPLOSION lake level greater than or equal to *9.0 ft. (Plant elevation)
Pump b a y level less t han -19.0 ft. Other events with sillllar hazard characteristics as determined by the Shift Manager or Emergency Director Loss or potential loss of any tw o b a rriers (Table F-1) [pg. 6 3) I UNPLA NNED lo u o f Control R oom Indication, to, 115 mlnutH orlo n ger with**lgniflc a n tt,.n ,t e nli n progreu (pg. 111) 1 ' SA3.1 An UNPLANNED even t results ln the I nab i lity to monitor one or more Table S-1 parameters from within t he Control Room for 15 rrinutes or longer ANO Any of the Table S-2 transient events are in progress Table 5-1 Safe ty Syst e m P aramete rs Readorpower RCS I Pressurizer Lever RCS I Pressurizer Pressure Core Exit I RCS Temperature Level in at least one steam generator Steam Generator Auxfliary Feed water Flow AutolY'l.l.
t lc or manual trip falls lo shut down !he re actor and 1ubseqoen 1 m a n.,al actlon1 t a kitn at the reactor control console* a,e no t , uccH1 ful ln s h utt ing do wn t he re actor [pg. 111] S A 6.1 Ar, automa ti c or manual trip fails to shut down the reactor as Indicated by reador power greater than 5% A NO Manual actions t aken at the reactor control consoles are not successful i n shutting down t he reactor. H nard ou , evenl aff ec ti ng a SAFETY SY STEM needed lo r t he curr e nt ope,.tlng mo de (po. 1 ZO] S A9.1 The occurrence of A N Y o f the Table S-3 hazardous even t s: ANO Event damage has caused Indications of degraded performance in one tra i n of a SAFETY SYSTEM needed for the current opera t i ng mode. ANO E ITHE R of the following:
* Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode , or
* The event has resulted In VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode. F A 1 2 Ally loss or any Potential loss of either Fuel C l ad or RCS barrier (Table F-1) [po. 63) U NPLANN ED Ion of Conlrol R oom lndlcallon, lor 15 mlnutH orlonger[pg10ill) 2 SU J.1 Ar, UNPLANNED event results In the In ab i lity to monitor one or mo r e of the Table S-1 parameters from within the Control Room for 15 minutes or longer. Reac t o r cool an t a c ti vity g, e ater th*n Technlcal Specillca tl on allo wa blellmit,[pg.101) 1 3 SU4.1 Failed Fuel Monitor 1 (2}-RE-109 reading greater than 750 mR/hr. S U 4.2 S~:f'~i t~~~l~nisa=~: t:~:: R~Jrn~~~ value Is ~pecifications as indicated by A~Y of the following conditions
: a. Dose Equivalent 1-131 greater than 50 &#xb5;Cifgm O R b. g:eo7~~:1~:;1Jo1~9~~a;:::~t~~5~iffrhi::~ess O R c.
greate r than 300 &#xb5;Ci/gm for RCS le*~oe tor 15 minute* or l or,ger [pg. 107] SUS.1 RCS unidentified or pressure boundary leakage greater than 10 gpm f or 15 minutes or longer. OR RCS Identified leakage greater than 25 gpm for 15 minutes or longer. OR Leakage from the RCS to a location outside containment.
or steam Generator tube leakage, greater than 25 gpm for 15 lllnutes or longer. AUlomatlc:
01 IY'l.l.nual trip falls to 1hu1 down lhe rtactor [pg 1091 SU6.1 Ar, automatic trip d i d not shut down the reactor ANO A subsequent manual act io n taken at the reactor control consoles is successful I n shutting down the reactor. SU6.2 A manual trip did not shutdown the reactor. ANO E I THER of the fol1o'Ning
:
* A su b sequent manual action taken at the r eactor control consoles is successful I n shutt i ng down t h e reactor O R I * :C,~~~:~::~~1ornatic trip is successful In shutting Lon o l all on1tte or offslte communications capabltl!IH {po. 111) S U7.1 Loss of ALL of the fo llowi ng onsite corrvnunication methods:
* Plant Pub i c Addtess System (Gal-Tronlcs)
Commercial PhonH
* PBX Phones
* Security Radio
* Portable Radios SU7.2 Loss of All o f the following otfsite response organization corrmunications methods: Nuclear Accidenl Reporting Syslem (NARS) Commerela l Phones
* PBX Phones
* Satenite Phones
* Manitowoc Coun t y Sherifr1 Department Radio S U 7.3 Loss of ALL of the follo'Ning NRC corrmunlcations methods;
* ns Phone System
* Commercial Phones
* PBX Phones
* Satelile Phones F aHUf* to l,o l a l e con ta i n ment or Ion ol containmen t prenur e con tr o l[pg.112] 2 S U 8.1 Fa ilur e of containment to isolate when required by an &ct\Jation s i gna l. ANO AL L required penetrations are not closed within 15 minu tes of the actuation signal. S U 8.2 Containment pressure greater than 25 psig ANO Less than one full tra in of Containment Cooling System equipment i s operating per design for 15 minutes or longer. Table F-1 Fission Product Barrier Matrix Fuel Clad (FC) Barrier jpg.68J Loss Conditions requiring entry I nto Core Cooling RED Path (CSP C.1) are met Potential Loss j Con d itions requiring en t ry into Co r e ! Cooling ORANGE Path (CSP C.2) a re !met /DR i Conditions requ i ring entry I nto Heat \S in k RED Path (CSP H.1) are mel Cont a inme nt radi a tion monitor reading j greate r th a n 577 Rlh r indicated on AN Y I of the following.
* 1(2)-RE-126 , 1(2)-RE-127
[
* 1(2)-RE-1 28 [ OR I 1(2)-RE-109 g,,a!e,thao 4 ,500 rr-R/lu i A NY con d ition In th e opinion of the Emer g ency Coordi n ator t hat I n dica t es Loss of the Fuel Clad b a rrier. A N Y condition In the opinion of the j Eme r gency Coo rd inator that indica t es !Po t e nt ial Loss of t he Fuel C l a d b a rri e r. Reactor Coolant System (RCS) Barrier jpg.68] Loss Ar, automatic or manual ECCS (SI) actuation required by E I THER of the following:
* UNISOlABLE RCS leakage OR
* SG t u b e RUPTURE Containment radiation monitor reading gre a ter th a n 1 1 R/hr indicated on ANY o f the f ollowing. 1(2)-R E-1 26 1(2)-RE-127
* 1(2)-RE-128 ANY condition In the opinion of the Emergency Coordinator that Indicates Loss o f the RCS ba r rier. Potential Loss Conditions requiring entry into Heat Sink RED Path (CSP H.1) are met OR Conditions requiring entsy in to RCS ln tegrtty RED Path (C SP P.1)are mel. Operation of a s t andby charg i ng (makeup) pu"1) is required by E ITH E R of the following:
* UNISOLABLE RCS leakage OR
* SG lube leakage A NY con d ition In the opinion o f the Emergency Coordinator that indicates Potential Loss of the RCS barrier. Containment (CNTMT) Barrier iPo 111 Loss A leaking or RUPTURED SG is FAUL TEO outside of containment Potent i al Loss Conditions req u iring entry I nto Core Cooling R E D Path (CSP C.1) are met. AN O CSP C 1 not effective wrth l n 15 minutes. Con t a in ment radiation monitor reading grea t er t h an 18 , 500 R/hr indicated on ANY of t he following. 1 (2}-RE-126 1(2}-RE-12 7
* 1 (2)-RE-128 Containment I solation is requ i red AND j Containment pressure greater than 60 El~~=n~:;~~~'::;:
: lty has been lost j 6: g based on Emergency Coordinator j 6% H 1 inside containment judgment ! O R OR l
* Containment pressure greater than
* UNISOlABLE pathway from J 25 psig contain ment to the environment exists i AN D O R j
* Less than one f\Jlt tra in of Indications of RCS leakage outside of ! ~~:=~~~~:::s~~t;~:
!nutes conta i nment ! or l onger. A NY condition ln t he opin i on of t he Emergency Coordinato r tha t in d icates Loss o f the Con t ainment barrier. ! ANY colldltlon in the opinion of the l Eme rgency Coordinator that indicates j Potentfal Loss o f the Containmen t r barrier. Modes: ~1~'~' -2~'~' -3~! ~' _4~i~I _5~''~ _G~!iDEF NEXT era* Point Beach Nuclear Plant EAL Classification Matrix Page 2 of 3 HOT CONDITIONS (RCS > 200'F) P o......er O pera tion S ta rtup H o t St an d by H o t S h u td o'Nfl Co ld Shutdown R e f u e li n g De f u ele d . 
' llllllllllllllllillltlllill'.lll GENERAL EMERGENCY I SITE AREA EMERGENCY I ALERT I UNUSUAL EVENT C Cdd lDI --. ._ -1 RCS Loni 2 Laoo ol Emorvenc1 ACPoww 3 RCS r...,. 4 Laoo ol 1111a1 DC -5 Lauol ....... 6 ......... ---' CG 1.1 Reactor ves seVR CS level cannot be monitored for JO rrinutes or longer. AND Core uncovery is indicated by ANY of the following: Containment High Radi at ion M onitors (1{2)-RE-1 26, RE-127 , or RE-128) reading greater than 100 R/hr Err a tic source range monitor i nd ication UNPLANNED incre ase In Containment Sum p A OR Waste Holdup Tank levels of sufficient magnitude to indicate core uncovery AND ANY indication from the Cont a inment Challenge Table C-1. CONTAINMENT CLOSURE not est a blished
* 6% H2 exists inside con ta i nment UNPLANNED i ncrease in containment pressure lf CONTAINMENT CLOSURE Is ,e .. sllbliahed prior to nceedi,g the >>minut e lime limit. theri deeleratiori ol a Geoer,IEm.rgencyls notraqulred. CS 1.1 Reactor vesseVRCS level cannot be monitored for 30 rrinutes or longer. AND Core uncovery Is I ndicated by ANY of the following: Containment High Rad i ation Monitor(1(2)-RE-126, RE-127, or RE-128) reading greater than 100 Rhlr Erratic source range monitor indication UNPLANNED increase in Containment Sump A OR Waste Holdup Tank levels of sufficient magnitude to In d icate co r e uncovery T-C-Z RCSHeol-apOu--
RCS Status Containment Heat-t1p Closure Status Duration Intact (bu t not N/A 60 min.* REDU CED INVENTORY)
Not intact Established 20 min.* OR REDUCED INVENTORY Not established Omln.
* If an RHR is i n operation within this t ime frame and RCS temperature is being re duced the EAL is not appllcable. **--* .. \;:.. ... Seismic even t (earthquake)
Internal or e xternal flooding event High winds o r tornado strike FI R E EXPLOSION Lake l evel greater than or equal to +9.0 ft {Pl ant e levation)
Pump bay level less than-19.0 fl Other events wit h similar hazard ch aracterist lcs a s determined by the S h ift Manager or Emergency Director Modes: 1 11 2 I I 3 I I 4 I I 5 I I 6 Power Operation Startu p Hot Stand!>; Hot Shutdown Cold S hu t d o'Ml Refuel i ng Loa1of R C81nY enlory l pn 48) 5
* CA1.1 Loss of reactor vesseVRCS inventory as indicated by level less than 16% on U-447 I U-447A. CA1.2 Reactor vesseURCS level canno t be monitored for 1 5 minutes or longer AND UNPLANNED increase in Containment Sump A OR waste Holdup Tank levels due to a loss of reactor vessel/RCS inventory. Lonofllll&deg;""'*and*l o ntlteAC,-lo~bl-.for gruterlha n 15"*""'-IPII &OJ DE F CA2.1 Loss of AL L offsite and ALL onsite AC Power to 1{2)-A05 and 1(2)-AOS for 15 minutes or longer. U NPI.ANNED Ion of RCS fell' 15 ffhllK or IDnfil., tp,, .W I 5
* CU1.1 UNPLANNED loss of reactor coolant results in reactor vesseV RCS level l ess than a required I01Ner limit for 15 minutes or longer. CU1.2 RCS water level cannot be monitored AND UNPLANNED increase in Containment Sump A OR waste Holdup Tank levels L oe* d .i 1 bul one AC,--* Mfeguai d buMe for 1 5 mnullleor'&deg;ntM
!P'l 41' CU2.1 DEF AC p<>Vt'er capability to 1(2)-A05 and 1{2~A06 is reduced to a single pOVt'er source f o r 15 minutes or lo ng e r AND Arty add iti onal single power source failure will result i n loss of all AC p01Ner to SAFETY SYSTEMS. IMbalytorNtirulnpial'lllncoldenutdawn iPO 5 1 1 UNPlANNE O lncl ... elnRC819mP9Jabe
{ 44J CA3.1 CUJ.1 UNPLANNED increase in RCS temperature to greater than UNPLANNED increase i n RCS temperature to greater 2oo*F f or greater t han the d uration specified i n Tab l e C-2. than 20D&deg;F CAJ.1 CUJ.2 UNPLANNED RCS p ress u re increase greater than 25 psig. Loss of ALL RCS temperature and reactor vessel/RCS (This EAL does not apply dur ing water-solid plant conditions.)
level ind i cation tor 15 minutes or longer. ewn1 affKlng 1 8AF&#xa3;TY SYSTEM fo r 1M -...ope,a11ngmNe
{Jl'J$JJ CA6.1 The occurrence of ANY of the Table C-3 hazardous events: AND 1 1 Event damage ha s caused indications of degraded performance in one train of a SAFETY SYSTEM n eeded for the current operating mode. AND EITHER of the following:
* Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode , or
* The event has resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current ope r ating mode NEXT era* DEF Defueled CU4.1 Indicated volt age is less than 11 5 VDC on required Vital DC buses 0-01 , 0-02, 0-03. 0-04 for 15 minutes or longer. DEF CU5.1 Loss of Al l of th e following o nsit e communication methods:
* Plant Public Address System (Gal-Tronlcs)
Commercial Phones
* PBX Phones
* Security Radio
* Portable Radios CU5.2 Loss of All of the f ollowing offsite response organization COl1'VT'llnicatio ns methods: Nuclear Acciderit Reporti n g System (NARS) Commercial Phones
* PBX Phones
* Satellite Phones
* Manitowoc County Sheriffs Department Rad io CU5.3 Loss of ALL of the following NRC communications methods:
* FTS Phone System Commercial Phones PBX Phones
* Satenite Phones Po i nt B ea ch Nuc l ear P l a n t EAL Cl as sifi ca tion M atrix P a g e 3 of 3 COLD CONDITIONS (RCS :S 200&deg;F) l}}

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