ML17346A766

Enclosures 2 & 3 to AEP-NRC-2017-56 - Response to Request for Additional Information Regarding the License Amendment Request to Revise Emergency Action Levels and EAL Technical Basis Manual

ML17346A766
Person / Time
Site:	Cook
Issue date:	12/08/2017
From:	Indiana Michigan Power Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML17346A762	List: AEP-NRC-2017-56, Response to Request for Additional Information Regarding the License Amendment Request to Revise Emergency Action Levels ML17346A766
References
AEP-NRC-2017-56
Download: ML17346A766 (250)

Text

Enclosure 2 to AEP-NRC-2017-56 Response to Request for Additional Information Regarding the License Amendment Request to Revise Emergency Action Levels to AEP-NRC-2017-56 Page 1 By letter dated May 23, 2017, (Reference 1), Indiana Michigan Power Company (l&M), the licensee for the Donald C. Cook Nuclear Plant* (CNP) Units 1 and 2, submitted a License Amendment Request. This amendment proposes to revise CNP Emergency Action Levels (EALs).

The Nuclear Regulatory Commission (NRG) staff in the Office of Nuclear Reactor Regulation is currently reviewing the submittal, and has determined that additional information is needed in order to complete the review (Reference 2). The text of the request for additional information (RAI) and l&M's response are provided below.

RAl-1 NE/ 99-01, Revision 6, Section 4.3, "Instrumentation Used for EALs," states, in part:

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 not use terms such as "off-scale low" or "off-scale high" since that type of reading may not be readily differentiated from an instrument failure.

The licensee's proposed EALs RS2.1 and RG2.1 Basis state that the setpoint for these EALs corresponds to the lower range for the spent fuel pool level instrument.

Please explain how this indication could not be confused with a failed instrument or provide a setpoint within the indicating range of the instrument.

l&M Response to RAl-1:

A reading of O feet (ft) on 1(2)-RLl-502-CRI Spent Fuel Pit Level Indication is a valid readable indication.

Upon receipt of the Spent Fuel Pool (SFP) low level alarm, operators will monitor and trend SFP level both remotely and locally. Such trending, as SFP level approaches the top of the fuel racks, reasonably excludes instrument failure for a reading of O ft. during a loss of SFP inventory event.

RAl-2 The licensee proposed EAL RA3.1 which includes the Central Alarm Station (GAS) and the Secondary Alarm Station (SAS) as threshold criteria. The Basis discussion states, "[t]he GAS and SAS are included in this EAL because of their importance to permitting access to areas required to assure safe plant operations ... "

• If the GAS and SAS can both provide access to areas required to assure safe plant operations, please consider revising EAL RA3. 1 to either provide an "AND" logic to the GAS and SAS or select the primary station as a threshold value.

to AEP-NRC-2017-56 Page 2 l&M Response to RAl-2:

The CAS is the primary station for plant access control. Therefore, SAS has been deleted from RA3.1 (See Enclosure 3, Page 54 of 236).

RAl-3 The licensee's proposed EAL RA3.2 includes Table R-2, "Safe Operation & Shutdown Rooms/Areas." This table appears to contain areas and/or mode applicability that would not prevent continued operation at power and would not prevent a plant shutdown and coo/down.

/ Please consider revising Table R-2 to reflect only the areas that require entry to either maintain operations in the current mode or to perform a plant shutdown or coo/down.

l&M Response to RAl-3:

CNP agrees with the RAI that the proposed Tables R-2/H-2 appear to include areas and modes not required to impact safe operation and shutdown to cold shutdown. CNP Operations Subject Matter Experts have re-evaluated the Attachment 3 analysis and have revised the mode dependent Tables H-2 and R-2 to reflect only those rooms/areas that require access to safely perform normal plant operation, shutdown and cooldown. Information has been added to the EAL Bases that supports the changes to Tables H-2 and R-2. (Enclosure 3, Page 55, 115, and 223 through 236 of 236)

RAl-4 For EALs RA3.2 and HA5.1, please address the following:

Table R-2/H-2, "Safe Operation & Shutdown Rooms/Areas," indicates that access is required to all levels of the Turbine Building. Additionally, Attachment 3, "Safe Operation &

Shutdown Rooms/Areas Table R-2/H-2 Bases," appears to include activities that, although desired, are not required for operation in the modes provided on Table R-2/H-2.

Please verify that Table R-2/H-2 include areas where access is required to support normal plant operations, coo/down or shutdown.

l&M Response to RAl-4:

See response to RAl-3 above.

RAl-5 The licensee states that CS1 example 1 and CS1 example 2 from NE/ 99-01, Revision 6, cannot be developed because no level indication exists that corresponds to the actual top of active fuel or 6 inches below the bottom inside diameter (ID) of the reactor coolant system (RCS) loop. A review of the current CNP EAL scheme indicates that CNP has an EAL CA 1. 1 threshold indication of Reactor Vessel Water Level < 614 ft. Please provide further justification for the removal of an EAL that relies on the reactor vessel level indicating to AEP-NRC-2017-56 Page 3 system (RVUS) from the proposed CNP EAL scheme, or revise accordingly. Note:

NE 99-01, Revision 6, developer notes provide guidance for indication that is "approximately the top of active fuel."

Please provide what RCS level indication is available near the bottom ID of the RCS loop and 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.

If there is not RCS level indication close to the top of active fuel, please explain why core uncovery based on a calculation was not included as a core uncovery indication for CNP CS1. 1 or revise accordingly.

l&M Response to RAl-5:

614 ft. on NLl-1000, RCS Half-Loop Wide Range Level Indication, corresponds to the minimum RCS level for Residual Heat Removal (RHR) pump operation. RCS level cannot be measured below 612 ft. on NLl-1000, which is below the bottom ID of the hot leg inlet. Should RCS level drop below this point it is assumed water level cannot be monitored.

Proposed CNP EAL CS1 .1 adequately addresses conditions when RCS level cannot be monitored

(< 612 ft. on NLl-1000).

RVLIS is not normally available or calibrated for use in Mode 5 and is disconnected in Mode 6 or Defueled. RVLIS is not required to be operable in Modes 5, 6 or Defueled. RVLIS may be used for trending; however, it is not calibrated for low temperature use.

The developer notes for CS1 and CG1 do not specify development of a calculation to determine RCS level in the absence of the ability to monitor level due to limitations of installed instrumentation.

It is not clear what inputs to such a calculation would be absent sump and tank level increases.

Generic CS1 example EAL_#3 (CNP EAL CS1.1) and CG1 example EAL #2 (CNP EAL CG1.1) provide the appropriate threshold criteria based on sump and tank level indications when RCS level cannot be monitored.

The EAL Technical Basis Manual has been revised to reflect these changes (See Enclosure 3, Page 65 of 236).

RAl-6 The licensee states that CG1 example 1 from NE/ 99-01, Revision 6, cannot be developed because no level indication exists that corresponds to 6 inches below the bottom ID of the RCS loop. Note: NE/ 99-01, Revision 6, developer notes for CG1 provide guidance for indication that is "approximately the top of active fuel" and not 6 inches below the bottom ID of the RCS loop.

Please provide further justification for the removal of an EAL that relies on the RVLIS from the proposed CNP EAL scheme, or revise accordingly.

to AEP-NRC-2017-56 Page4 If there is not RCS level indication close to the top of active fuel, please explain why core uncovery based on a calculation was not included as a core uncovery indication for CNP CG 1. 1 or revise accordingly.

l&M Response to RAl-6:

See the response to RAl-5 above regarding RVLIS and core uncovery calculation.

RAl-7 The CNP EAL CU3. 1 Basis discussions states, "[i]n the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on time to boil date ... "

The CNP EAL CA3. 1 Basis discussion states, "[i]n the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RCS pressure increase criteria when in Mode 5 or based on time to boil data when in Mode 6 ... "

Please provide clarification as to whether or not CNP will Jose the ability to monitor RCS temperature when decay heat removal capability is lost. If so, please add the time to boil criteria to the EAL threshold value. If not, please consider revising the Basis discussion as necessary to reflect CNP available indications for RCS temperature indications.

l&M Response to RAl-7:

Under normal conditions, while in Mode 5, a loss of RHR flow (decay heat removal capability) would not result in a loss of the ability to monitor RCS temperature. In Mode 5, lncore Thermocouples are available to detect valid RCS temperatures independent of decay heat removal capability. In Mode 6, with the reactor vessel head removed a loss of RHR could result in a loss of valid RCS temperature indications unless temporary thermocouples are installed in the RCS. The cited bases guidance provides alternative means of indirectly assessing RCS temperature relative to the 200°F threshold The bases for both CU3.1 and CA3.1 provide the following description of available RCS temperature indications:

• NTl-100, NT/-101, Selected lncore Temperature or Temporary Thermocouples

• NTR-210, Reactor Coolant T-Cold Wide Range Loop 1

• NTR-220, Reactor Coolant T-Cold Wide Range Loop 2

• NTR_-230, Reactor Coolant T-Cold Wide Range Loop 3

• NTR-240, Reactor Coolant T-Cold Wide Range Loop 4

• NTR-110, Reactor Coolant T-Hot Loop 1

• NTR-120, Reactor Coolant T-Hot Loop 2

• NTR-130, Reactor Coolant T-Hot Loop 3

• NTR-140, Reactor Coolant T-Hot Loop 4

• RHR display on PPG to AEP-NRC-2017-56 Pages Additionally, the following was added to the CU3.1 and CA3.1 bases:

"Refer to OHP-4022-017-001, Loss of RHR Cooling."(See Enclosure 3, Page 80, 81, 85, and 86 of 236)

RAl-8 The second paragraph of CNP EAL HU2. 1 Basis discussion includes a discussion that provides direction that implies that CNP should obtain offsite verification to "avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity." This paragraph does not contain wording that this verification does not preclude a timely emergency verification.

Although the guidance relative to timely emergency action declaration is contained in the fourth paragraph of the Basis discussion, there is a potential that a decision maker may take actions as described in the second paragraph and not see the appropriate wording.

Additionally, the provided CNP threshold value does not rely on actuation of seismic instrumentation.

The fourth paragraph contains a lateral acceleration value that does not seem consistent with CNP operating basis earthquake criteria. This could potentially imply that the fourth paragraph contains generic information that may not specifically apply to CNP.

The current CNP EAL scheme includes seismic instrumentation activation as a threshold value.

a. Please explain, in greater detail, why the CNP seismic instrumentation cannot be used to determine an actual or potential seismic event, or revise accordingly.

b. Please revise the Basis discussion for HU2. 1 to clarify the actual EAL criteria and to ensure that guidance relative to timely emergency declaration is readily available to

• decision makers.

l&M Response to RAl-8 The cited fourth paragraph regarding event verification with external sources has been moved to be the first basis paragraph.

The cited generic Operating Basis Earthquake (QBE) ground acceleration (0.08g) has been revised to reflect the CNP specific QBE ground acceleration (0.1 Og) (See Enclosure 3, Page 101 and 102 of 236).

• As stated in the basis:

Procedure 1(2)-0HP-4022-001-007 Earthquake provides the guidance for determining if the OBE earthquake threshold is exceeded and any required response actions. (ref. 2). Because CNP seismic instrumentation does not provide direct and timely indications of having exceeded the OBE

.Enclosure 2 to AEP-NRC-2017-56 Page 6 ground acceleration, the alternative EAL wording specified in the generic NE/ 99-01 HU2 developers note (felt earthquake) is implemented.

The CNP seismic instrumentation cannot be used to determine an actual or potential seismic event greater than OBE because the data that is recorded and processed by the Seismic Monitoring System cannot be reviewed and analyzed to determine the acceleration value within the 15 minutes required.

RAl-9 The licensee's proposed EAL HU3.3 Basis states, "[a]s used here the term 'offsite' is meant to be areas external to the CNP plant PROTECTED AREA." That statement infers that a different definition is normally used for the term "offsite."

Please consider including that statement in the EAL and on the wallboard to prevent possible misclassification, or provide a justification as to why no statement is necessary.

l&M Response to RAl-9:

EAL HU3.3 has been revised to read as indicated below and the basis statement was deleted (See Enclosure 3, Page 106 of 236):

"Movement of personnel within the plant PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)"

RAl-10 Concerning CNP EAL HU4.2, Table H-1 "Fire Areas" includes the containment. This could result in an event declaration due to the spurious actuation of a single fire alarm.

Based on the information provided in the LAR, the staff could not determine if the containment fire detection system at CNP, in combination with the CNP containment ventilation system supported the inclusion of the containment as a fire area for EAL HU4.2.

Please justify including the containment as a Fire Area for HU4. 2, or revise accordingly.

l&M Response to RAl-1 O:

The fire detection system in the CNP containments consists of thermistor strings in cable trays. This type of detection system is not typically subject to spurious alarms. CNP Operators rely on

. containment temperature (indicated in the Control Room) as an indirect method of fire detection when a thermistor string is removed from service. An analysis based on air mixing in containment would not be appropriate for CNP as the thermistor strings detect heat and not particulate (smoke) in the containment atmosphere.

Enclosure 2 to AEP-NRC-2017-56 Page 7 RAl-11 EAL HA6. 1 states the following in the Basis discussion:

Plant control is considered to have been transferred when either 1) control of the plant is no longer maintained in the Control Room or 2) the last licensed operator has left the Control Room, whichever comes first.

This EAL is only applicable when the decision has been made to evacuate the Control Room, not when conditions are being evaluated.

EAL HS6. 1 states the following in the Basis discussion:

The 15-minute time period starts when either 1) control of the plant is no longer maintained in the Control Room or 2) the last licensed operator has left the Control Room, whichever comes first.

The first paragraph under EAL HA6. 1 above, implies that as soon as control of the plant is no longer maintained in the Control Room that "[p]lant control is considered to. have been transferred" or "the last operator has left the Control Room." Considering that no one may have actually left the Control Room at the time that plant control is lost, and that no one may actually be at the alternate shutdown location, this paragraph could lead to inconsistent EAL declarations.

The second paragraph under EAL HA6. 1 above, implies that the EAL start time is when the decision has been made to evacuate the Control Room. This statement could imply that loss of plant control from the Control Room would only apply once the decision is made to evacuate the Control Room.

The language in HA6. 1 may lead to inconsistent implementation of HS6. 1. Please revise the Basis discussions for HA6. 1 and HS6. 1 to provide a clear 15 minute start time to support consistent EAL declarations.

l&M Response to RAl-11:

The first and third paragraphs of the HA6.1 basis have been deleted (See Enclosure 3, Page 118 of "236):

"Pia ht control is considered to have been transferred when either 1) control of the plant is no longer maintained in the Control Room or 2) the last licensed operator has left the Control Room, whichever comes first" "This EAL is only applicable when the decision has been made to evacuate the Control Room, not 1.vhen conditions are being evaluated."

to AEP-NRC-2017-56 Pages The first paragraph of the HS6.1 basis has been deleted (See Enclosure 3, Page 120 of 236).

"The 15 minute time period starts when either 1) control of the plant is no longer maintained in the Control Room or 2) the last licensed operator has left the Control Room, whichever comes fifSt.

RAl-12 For the proposed CNP EALs SU6.1; SA6.1 and SS6.1 a power level (greater than or equal to 5%) was added to the EALs. The intent of NE/ 99-01, Revision 6 is to align the EAL classifications with site-specific emergency operating procedure (EOP) criteria for a successful reactor shutdown, which would benefit the decision makers by providing consistent criteria. The power level provided in the developer notes in NE/ 99-01, Revision 6 is an example that represents one typical EOP indication for a generic power plant and was not intended to be a complete list of EOP indications for any specific power plant. The Reactor Trip or Safety Injection* Action/Expected Response for typical Westinghouse plants include reactor power decreasing and rod insertion indications to determine if the anticipated transient without scram (A TWS) response procedure should be initiated. Additionally, the subcriticality critical safety function status trees for typical Westinghouse plants include both power and startup rate to determine if an A TWS has occurred. The proposed A TWS EALs for CNP appear to rely solely on a reactor power of 5% as an indication of a reactor trip.

Please clarify CNP EALs SU6.1, SA6.1 and SS6.1 to reflect the EOP reactor shutdown criteria in the EOPs, or describe how the wording similar to NE/ 99-01, Revision 6 will be used to support timely and accurate assessment of CNP EALs SU6.1, SA6.1 and SS6.1.

l&M Response to RAl-12:

For CNP 5 percent(%) reactor power is the site-specific power level specified in the CNP EOPs for reactor shutdown and corresponds to shutdown decay heat levels (ensures that the heat load to available heat sinks is just the decay heat level normally accommodated with Auxiliary Feedwater flow).

Per the generic NEI 99-01 Revision 6 guidance, EALs SU6.1, SU6.2, SA6.1 and SS6.1 are only applicable in Mode 1, which corresponds to a condition in which reactor power is ~ S%. The generic guidance bases recognizes that the concern, relative to reactor shutdown, is reactor power greater than decay heat levels following a reactor trip signal (manual or automatic).

The NEI 99-01 Revision 6 SUS, SAS and SSS Developer Notes state:

"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)."

CNP has evaluated the EOP guidance and determined that reactor power level is the best indication, which is consistent with industry benchmarking.

to AEP-NRC-2017-56 Page 9 RAl-13 Concerning CNP EAL SS6.1, the Basis discussion includes guidance that emergency boration or manually driving control rods are credited as a successful manual trip. This is not consistent with the successful manual trip discussion contained in the Basis discussion for SA6. 1.

• Additionally, the Basis discussion for SS6. 1 seems to be intended to address the .

"[a]II actions to shut down the reactor are not successful ... " threshold value.

Please revise the SS6. 1 Basis discussion to clarify the inability to shut down the reactor as a requirement for declaration of SS6. 1 and to prevent a conflict with the definition of a manual trip provided by SA6. 1 or explain how the proposed SS6.1 basis wording will not result in inconsistent EAL declarations.

l&M Response to RA1.:13:

The cited basis has been revised to read (See Enclosure 3, Page 161 of 236):.

"Reactor shutdown achieved by use of FR-S. 1 Response to Nuclear Power Generation/A TWS such as tripping the main turbine, locally opening reactor trip breakers, emergency boration or manually driving control rods are also credited as a successful means of shutting down the reactor provided reactor power can be reduced.be/ow 5% before indications of an extreme challenge to either core cooling or heat removal exist" .

RAl-14 For initiating condition (IC) Reactor Coolant System Fission Product Barrier A.2, Potential Loss, the staff could not specifically determine which critical safety function status tree Integrity-RED Path was intended by this IC. Please provide clarification as to the specific RED path that should be used to assess this condition.

• l&M Response to RAl-14:

CNP has only one defined Critical Safety Function Status Tree (CSFST) Integrity RED Path (F-D.4) as specified in RCS Fission Product Barrier Potential Loss A.2 and 1(2)-0HP-4023-F-0.4:

"CSFST Integrity-REP Path (F-0.4) conditions met" The words "Reactor Coolant" and "RCS" from the first basis paragraph have been deleted to align with the proper CNP CSFST title (See Enclosure 3, Page 194 of 236).

RAl-15

_ The proposecj QNP EAL Technical Basis sections tor the loss. and potenti<J/ loss of the Fuel Clad, RCS, and Containment barriers includes the_ following:

The SEC [site emergency coordinator] judgment threshold addresses any other. factors relevant to determining if the [Primary Containment/RCS/Fuel Clad] barrier is lost. Such a determination should include IMMINENT barrier degradation, barrier monitoring capability and dominant accident sequences. .

to AEP-NRC-2017-56 Page 10

• Imminent barrier degradation exists if the degradation will likely occur within relatively short period of time based on a projection of current safety system performance. The term "imminent refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability. concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the-' f=.OPs. The SEC should be mindful of the Loss of AC power (Station Blackout) and A 7WS EALs to assure timely emergency classification declarations.

The proposed CNP basis wording appears to bound and/or modify the fission product barrier thresholds which rely on the opinion gf the SEC for an indication of loss or potential loss of fission product barriers. Please clarify the proposed CNP EAL Technical Basis section for the loss and potential loss of the Fuel Clad, RCS, and Primary Containment barriers as necessary *to _remove any wording that could either bound and/or modify the

_judgement of the SEC concerning a loss, or potential loss, of a fission product barrier, or explain how this wording will not potentially inhibit the SEC's judgement.

l&M Response to RAl-15:

The cited bases paragraphs have been deleted from the Fuel Clad, RCS and Containment SEC judgment bases (See Enclosure 3, Page 190, 191, 202, 203, 221, and 222 of 236).

RAl-16 Co_ncerning CNP EALs CA 6.1 and SA9.1, based on discussions between the NRG and NE/1/ndustry, the staff provided a response to Emergency Preparedness Frequently Asked

.* Guestion (EPFAQ) 2016-002, "Clarification of Equipment Damage as a Result of Hazardous Event" Although the current guidance provided by NE/_ 99-01, Revision 6, remains an acceptable method to develop these EALs, the clarifications provided by EPFAQ2016-002 can be used to ensure consistent EAL classifications that are aligned with the intent of an Alert classification.

Please consider implementing the NRG proposed solution for EPFAQ 2016-002 for CNP EALs CA6.1 and SA9.1. ..

l&M-Response to RAl-16:

ICs CA6 and SA9 and associated EALs CA6.1 and SA9.1 have been revised to be consistent with EPFAQ 2016-002, as follows (See Enclosure 3, Page 91, 92, 93, 167, 168, and 169 of 236).

to AEP-NRC-2017-56 Page 11 "Hazardous event affecting SAFf=TY SYSTEMS needed for the current operating mode" "The occurrence of any Table C-6 hazardous event AND

• Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:*

• Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode

• Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 11, 12)"

The word "a" has been revised to read "the" in both of the bulleted thresholds to clarify that the intent of the EAL is degraded performance or visible damage to the second train of the same safety system consistent with the referenced EPFAQ bases.

New notes 11 and 12 have been added as follows:

"Note 11:/f the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 12:/f 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."

CA6.1 and SA9.1 bases and definition of VISIBLE DAMAGE has been revised to support the revised EAL wording consistent with EPFAQ 2016-002.

Non RAI Changes:

General minor typographical and format corrections have been made to the EAL Technical Basis Manual. These changes are not associated with RAI responses:

to AEP-NRC-2017-56 Page 12

References:

1. Letter from Q. S. Lies, Indiana Michigan Power Company (l&M), to U. S. Nuclear Regulatory Commission (NRC), "Donald C. Cook Nuclear Plant, Unit 1 and Unit 2, License Amendment Request to Revise Emergency Action Levels," dated May 23, 2017. Agency wide Documents Access and Management System Accession (ADAMS) No. ML17146A073.

2. Email from J. Rankin, NRC, to H. L. Kish, l&M, Donald C. Cook Nuclear Plant -Requests For Additional Information - License Amendment Request - Emergency Action Level Scheme Change," dated October 25, 2017. ADAMS Accession No. ML17298C053.

Enclosure 3 to AEP-NRC-2017-56 EAL Technical Basis Manual (Revision 0)

. AEP: D.C. Cook EAL Technical Basis Manual Revision 0 RAI Response Final Clean 12/6/17 Page 1 of 236 INFORMATION USE

TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE ..............................................................................................................................3 2.0 DISCUSSION ................................... .-.................................. :........................... :............. :......... 3 2.1 Background ..........................................................................................................................3

-2.2 - Fission Product Barriers; .......-.............-....... ,...... ,...... ,...... ,....-.. -....... -.....-.. -.....-.. -.....................-.... .-4 2.3 Fission Product Barrier Classification Criteria ..................................................................... .4 2.4 EAL Organization .................................................................................................................5 2.5 Technical Bases Information ................................................................................................6 2.6 Operating Mode Applicability .... :...... :...... :............................................................................8 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ............................................... 9 3.1 General Considerations .......................................................................................................9

• 3.2- Classification Methodology ........ :...... :...... :...... :...... :...... :...... :...... :...... : ...... : ...... : ............. -.... 1O

4.0 REFERENCES

.....................................................................................................................13 4.1 Developmental ...................................................................................................................13 4.2 Implementing .....................................................................................................................13 5.0 - DEINITIONS, ACRONYMS & ABBREVIATIONS .................................................................. 14 6.0 Cook TO NEI 99-01 Rev. 6 EAL CROSS-REFERENCE ....................................................... 22 7.0 ATTACHMENTS ...................................................................................................................26 1 Emergency Action Level Technical Bases ................................................................ 27 Category R Abnormal Rad Release/ Rad Effluent.. ........................................ 27 Category E ISFSI ............................................................................................57 Category C Cold Shutdown / Refueling System Malfunction ........................... 60 Category H Hazards ................. : ...................................................................... 94 Category S System Malfunction .................................................................... 130 Category F Fission Product Barrier Degradation .......................................... 170 2 Fission Product Barrier Loss/ Potential Loss Matrix and Bases ....................................................................................................175 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases ..................... 223 Page 2 of 236 INFORMATION USE

1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for D. C. Cook Nuclear Plant (CNP). Decision-makers responsible for implementation of PMP-2080-EPP-101 Emergency Classification, may use this document as a technical reference in support of EAL interpretation. This information may assist the SITE EMERGENCY COORDINATOR (SEC) in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizabl_e for the classification, but. within 15 minutes or less in an cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.

Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Coordinator refers to it during an event), the NRG staff expects that changes to the basis document will be evaluated* in accordance with the provisions of 10 CFR 50.54(q). Additionally, changes to plant AOPs and EOPs that may impact EAL bases shall be evaluated in accordance with the provisions of 10 CFR 50.54(q).

2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the CNP Emergency Plan.

In 1992, the NRG endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," EAL guidance.

NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:

• Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.

• Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSls).

• Simplifying the fission product barrier EAL threshold for a SITE AREA EMERGENCY.

Subsequently, Revision 6 of NEI 99-01 "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ADAMS Accession Number ML12326A805) (ref. 4.1.1) was issued which incorporates resolutions to numerous implementation issues including the NRG EAL Frequently Asked Questions (FAQs). Using NEI 99-01 Revision 6, CNP conducted an EAL implementation upgrade project that produced

• the EALs discussed herein.

Page 3 of 236 INFORMATION USE

2.2 Fission Product Barriers 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.

Many of the EALs derived from the NEI methodology are fission product barrier threshold

• based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of.the three fission product barriers. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss" threshold means the barrier no longer assures containment of radioactive materials. A

".Potential Loss" threshold implies an increased probability of barrier loss and decreased .

certainty of maintaining the barrier.

The primary fission product barriers are:

A Fuel Clad.(FCj: The Fuel Ciad Barrier consists ofthe cladding material that contains the

• fuel pellets.

• B. Reactor Coolant System (RCS): 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.

C. Containment (CNMT): 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 Emergency Classification Level (ECL) from ALERT to a SITE AREA EMERGENCY or a GENERAL EMERGENCY.

2.3 Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss:

Alert:

Any loss or any potential loss of either Fuel Clad or RCS barrier Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

Loss of any two barriers and loss or potential loss of the third barrier Page 4 of 236 INFORMATION USE I

2.4 EAL Organization The CNP EAL scheme includes the following features:

• Division of the EAL set into three broad groups:

o EALs applicable under any plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered.

• o

• EALs applicable only under hot operating modes - This group would only be

. reviewed by the EAL-user when the plant is in Hot Shutdown, Hot Standby, Startup, or Power Operation mode.

o . EALs applicable only under cold operating modes -This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or.

Defueled mode.

The purpose of the groups is to avoid review of hot condition EALs when the plant is in a

a cold *condition and avoid review of cold condition .EALs when the plant is in hot condition. This approach significantly minimizes the total number of EALs that must be **

reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.

• Within each group, assignment of EALs to categories and subcategories:

Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The CNP EAL categories are aligned to and represent the NEI 99-01" Recognition Categories." Subcategories are used in the CNP EAL scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds.

Page 5 of 236 INFORMATION USE

EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory I

Any Operating Mode:

R - Abnormal Rad Levels / Rad Effluent 1 -:- Radiological Effluen.t 2.:.... Irradiated Fuel Event 3 - Area Radiation Levels .

H - Hazards and Other Conditions 1 - Security Affecting Plant Safety 2 - Seismic Event 3- Natural or Technological Hazard 4- Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7:.... Em*ergericy Coordinator Judgment*

E- ISFSI

• 1 - Confinement Boundary Hot Conditions:

S - System Malfunction 1 - Loss of Emergency AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4 - RCS Activity 5 - RCS Leakage 6- RPS Failure 7 - Loss of Communications 8 - Containment Failure 9 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions:

C - Cold Shutdown / Refueling System 1-RCS Level Malfunction 2 - Loss of Emergency AC Power 3- RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL Technical Bases Document in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information.

2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (i.e.

Any, Hot, Cold), EAL category (i.e. R, C, H, S, E and F) and EAL subcategory. Where applicable. A summary explanation of each category and subcategory is given at the I Page 6 of 236 INFORMATION USE *1

beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:

Category Letter & Title Subcategory Number & Title Initiating Condition (IC)

Site-specific description of the generic IC given in NEI 99-01 Rev. 6.-

EAL Identifier (enclosed in rectangle)

Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier: *

1. First character (letter): Corresponds to the EAL category as described above (R, C, H, S, E or F) .

2, Second character (letter): The emergency classification {G, S, A or U)

G = General Emergency S = Site Area Emergency A= Alert U = Unusual Event

3. Third character (number): Subcategory number within the given category.

Subcategories are sequentially numbered beginning with the number one (1 ). If a category does not have a subcategory, this character is assigned the number one (1 ).

4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).

Classification (enclosed in rectangle):

Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)

EAL (enclosed in rectangle)

Exact wording of the EAL as it appears in the EAL Classification Matrix Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 -

Cold Shutdown, 6 - Refueling, D - Defueled, or Any. (See Section 2.6 for operating mode definitions)

Definitions:

If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1.

Basis:

A basis section that provides CNP-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.

Page 7 of 236 INFORMATION USE

CNP Basis Reference(s):

Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (ref. 4.1. 7) 1 Power Operation Keff ~ 0.99 and reactor thermal power> 5%

2 .Startup ..

Keff ~ 0.99 and reactor thermal power :5 5% .

3 Hot Standby Keff < 0.99 and average coolant temperature ~ 350°F 4 Hot Shutdown .

Keff < 0.99 and average coolant temperature 350°F > T avg > 200 5 Cold Shutdown Keff < 0.99 and average coolant tem*perature*s 200°F

• 6 Refueling **

One or more reactor vessel head closure bolts are less than fully tensioned D Defueled All reactor fuel removed from reactor pressure vessel (full core off load during refueling or extended outage).

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 wa_s 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 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.

Page 8 of 236 INFORMATION USE ,.

3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Coordinator 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 based on loss or potential loss of Fission Product Barrier Thresholds.**

3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition Withiri 15 minutes afterthe 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, E:mergency Planning for Nuclear Power Plants" (ref.4.1.10). **

When assessing an EAL 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."

3.1.2 VALID Indications All emergency classification assessments shall 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, verification 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.

An indication, report, or condition is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Coordinator 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.

3.1.4 Planned vs. UNPLANNED Events 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 Page 9 of 236 INFORMATION USE

• 1

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 (ref.

4.1.4).

3.1.5 Classification Based on Analysis 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.). For these EALs, the EAL wording 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 NRG 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).

3.1.6 SITE EMERGENCY COORDINATOR (SEC) Judgment 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. The NEI 99-01 EAL scheme provides the SEC with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (EGL) definitions (refer to Category H). The SEC will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular EGL definition.

A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.

3.2 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 must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the EGL must be declared in accordance with plant procedures no later than fifteen minutes after the process "clock" started.

When assessing an EAL 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/DPR-ISG-01 (ref. 4.1.10).

3.2.1 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:

i- * - If two ALERT EALs are met, whether at one unit or at two different units, an ALERT should be declared.

Page 10 of 236 INFORMATION USE

Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRG Guidance for Emergency Notifications During Quickly Changing Events (ref. 4.1.2).

3.2.2 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 beforethe 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 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.

3.2.3 Classification of IMMINENT Conditions Although EALs provide specific thresholds, the Emergency Coordinator 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 Coordinator, 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 implementation of protective measures.

3.2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements 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.

As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 4.1.2).

3.2.5 Classification of Short-Lived Events Event-based ICs 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 an earthquake or a failure of the reactor protection system to automatically trip the reactor followed by a successful manual trip.

3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs 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 Page 11 of 236 INFORMATION USE

,_._ _J

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 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 riot required.- Fm illustrative purposes, consider the following example:

An ATWS occurs and the high pressure ECCS systems fail to automatically start. RPV

• level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the fuel clad and RCS barriers) .. If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only.

It is important to stress that the 15-minute emergency classification assessment period (process clock) 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 when an operator is able to take a successful corrective action prior to the Emergency Coordinator 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.

3.2.7 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 personnel 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, Event Report Guidelines 10 CFR 50.72 and 50.73, (ref. 4.1.3) is applicable.

Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50. 72 (ref.

4.1.4) 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.

3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 4.1.3).

Page 12 of 236 INFORMATION USE . ,

4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 4.1.2 RIS 2007-02 Clarification of NRG Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007.

4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.7;3 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors .

4:1.5 1O § CFR 50.73 License Event Report System 4.1.6 CNP UFSAR Figure 1.3-1 Plot Plan 4.1.7 Technical Specifications Table 1.1-1 Modes*

4:1.8 PMP-41 oo.:sDR-001 Pla*nt Shutdown Safety and Risk Management .

4.1.9 PMP-201 O-PRC-001 Procedure Writing 4.1.10 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.11 CNP Emergency Plan 4.2 Implementing 4.2.1 PMP-2080-EPP-101 Emergency Classification 4.2.2 NEI 99-01 Rev. 6 to CNP EAL Comparison Matrix 4.2.3 CNP EAL Matrix Page 13 of 236 INFORMATION USE

5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (ref. 4.1.1 except as noted)

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.

ALERT Events are jn progress, or have occurred, which involv~ an actual or pote_ntial.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 small fractions of the EPA Protective Action Guideline expo"sure levels.

CONTAINMENT CLOSURE

• The procedurally defined *actions takeri to s*ecure co"ntairimerit and its *associated structures; systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to CNP, Containment Closure is established when the requirements of PMP-4100-SDR-001 are met (ref. 4.1.8).

CONFINEMENT BOUNDARY The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As applied to the CNP ISFSI, the CONFINEMENT BOUNDARY is defined to be the Multi-Purpose Canister (MPC).

EMERGENCY ACTION LEVEL 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 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) and GENERAL EMERGENCY (GE).

EPAPAGs Environment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsite exposures in excess of the EPA PAGs requires CNP to recommend protective actions for the general public to offsite planning agencies.

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 require a post-event inspection to determine if the attributes of an explosion are present.

Page 14 of 236 INFORMATION USE

• 1

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 driye 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.

FISSION PRODUCT BARRIER THRESHOLD A pre.,.determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.

FLOODING A condition where water is entering a room or area faster than installed equipment is capable a

of removal: resulting in rise of water level within the room.or area.. . .. .. . . ...

GENERAL EMERGENCY 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 ACTIONS that result 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.

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 CNP 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 CNP. 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.

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.

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IMPEDE(D)

Personnel access to a room or area is hindered to an extent that 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).

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)

_A complex that is_ designed and constructed for t_he ir)terim_storage of spent nucl~ar_fuel and __

other radioactive materials associated with spent fuel storage.

Initiating Condition 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.

INTACT (RCS)

The RCS should be considered intact when the RCS pressure boundary is in its normal condition fo_r the col~ shutdown mode_of operation (e.g.~ no free:z:e se.als or no_zzle da111s).

MAINTAIN Take appropriate action to hold the value of an identified parameter within specified limits.

OWNER CONTROLLED AREA The property associated with the station and owned by the company. Access is normally limited to persons entering for official business.

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 encompassed by physical barriers to control access to the plant and to the ISFSI.

(ref. 4.1.6).

REDUCED INVENTORY Operating condition when fuel is in the reactor vessel and Reactor Coolant System level is lower than 3 feet (or more) below the Reactor Vessel flange (ref. 4.1.8).

REFUELING PATHWAY The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway.

Page 16 of 236 INFORMATION USE

RUPTURED The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.

RESTORE Take the appropriate action required to return the value of an identified parameter to the applicable limits SAFl=TY 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 classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure ~oundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

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 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 ACTIONS that result 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 Guidelines exposure levels beyond the site boundary.

SITE EMERGENCY COORDINATOR (SEC)

The individual who has the responsibility for event classification, event notification, and approval of protective action recommendations to offsite organizations. It is recognized that during an emergency event this responsibility can be formally turned over from the Shift Manager, to a Site Emergency Coordinator located in the TSC, or to the Emergency Director located in the EOF as the response facilities become activated during an emergency event.

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.*

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UNUSUAL EVENT Events are in progress or have occurred which indicate a potential degradation in 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.

VALID An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need.

• for timely assessment.

VISIBLE DAMAGE Damage to a component or structure that is readily observable without measurements, testing,.

pr analy?is. The visual impact of the damage is ?uffic.ientjo cctuse.concem regctrding th~

operability or reliability of the affected component or structure.

Page 18 of 236 INFORMATION USE

5 .2 Abbreviations/Acronyms

~F ....................................................................................................... Degrees Fahrenheit 0

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • Degrees AC ........................ _. ............. _. ..........................._. ......_. ......_. ............ '. ...... '. .. Alt.ernating .Current .

ATWS ...................................................................... Anticipated Transient Without Scram COE ****.******'.*************:******:******:******:******:*,****:*,****.-*,******., ...... ,.... Committed Dose .Equivalent CF.R ****.*******:*****'.******'.******'.********************:********************:******.***** Code of Federal R.egulations CHRM ........................................................................... Containment High Range Monitor CNMT ............................................................................................................Containment CNP ........................................................................................... D. C. Cook Nuclear Plant.

CSFST ....................................................................... Critical Safety Function Status Tree D ..........................................................................................................................Defueled OBA ............................................................................................... Design Basis Accident DBT .: .................... *...... ~*.................................................. :...... :.. :... :..... ::oes.ign Basis Threat ..

DC ............................................................................................................... Direct Current EAL ............................................................................................. Emergency Action Level ECCS ............................................................................ Emergency Core Cooling System ECL ................................................................................. Emergency Classification Level EOF .................................................................................. Emergency Operations Facility EOP ............................................................................... Emergency Operating Procedure EPA .............................................................................. Environmental Protection Agency ERG ................................................................................ Emergency Response Guideline EPIP ................................................................ Emergency Plan Implementing Procedure ESF ......................................................................................... Engineered Safety Feature ESW ........................................................................................ Emergency Service Water FAA .................................................................................. Federal Aviation Administration FBI ................................................................................... Federal Bureau of Investigation FEMA. .............................................................. Federal Emergency Management Agency FPB ................................................................................................ Fission Product Barrier FSAR .................................................................................... Final Safety Analysis Report GE ..................................................................................................... General Emergency Hrs ........................................................................................................................... Hours IC ......................................................................................................... Initiating Condition IPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI. ........................................................... Independent Spent Fuel Storage Installation Kett ......................................................................... Effective Neutron Multiplication Factor LCO .................................................................................. Limiting Condition of Operation LER ............................................................................................... Licensee Event Report LOCA .... :...... : ...... :................... :......................................... :........ Loss of Coolant Accident LWR ................................................................................................... Light Water Reactor Page 19 of 236 INFORMATION USE

MPG ................................... Maximum Permissible Concentration/Multi-Purpose Canister mR, mRem, mrem, mREM .............................................. mil Ii-Roentgen Equivalent Man MSL ........................................................................................................ Main Steam Line NEI .............................................................................................. Nuclear Energy Institute NESP ................................................................... National Environmental Studies Project NPP .................................................................................................. Nuclear Power Plant NRC ................................................................................ Nuclear Regulatory Commission

• NSSS ........................ :........................... :............. :...... :...... Nuclear Steam Supply System NORAD ................................................... North American Aerospace Defense Command (NO)UE ............ ,...... ,...... ,...... ,...... ,...... ,...... ,.......................... Notification of Unusual Event NU MARC ........................................... Nuclear Utility Management and Resource Council OBE ...................................................................................... Operating Basis Earthquake OCA: ...... :...... : ............. : ..... .-...... .-...... .-.............................................. Owner Controlled Area*

ODCM: ...... :..... :: ...... :.. ~ .......... ~ ...... ~ ..... :...... :...... :............. Off-site Dose-Calculation Manual ORO ................................................................................. Offsite Response Organization PA .............................................................................................................. Protected Area PAG ........................................................................................ Protective Action Guideline PRA/PSA ..................... Probabilistic Risk Assessment/ Probabilistic Safety Assessment PWR ....................................................................................... Pressurized Water Reactor PSIG ................................................................................ Pounds per Square Inch Gauge R ........................................................................................................................ Roentgen RCC ............................................................................................ Reactor Control Console RCS ............................................................................................ Reactor Coolant System Rem, rem, REM ....................................................................... Roentgen Equivalent Man RETS ......................................................... Radiological Effluent Technical Specifications RPS ........................................................................................ Reactor Protection System R(P)V ....................................................................................... Reactor (Pressure) Vessel RVLIS ................................................................. Reactor Vessel Level Indicating System SID .................................................................................................................... Shutdown SAR ............................................................................................... Safety Analysis Report SBO ......................................................................................................... Station Blackout SCBA ....................................................................... Self-Contained Breathing Apparatus SEC ...................................................................................... Site Emergency Coordinator SG ......................................................................................................... Steam Generator SI .... *.......................................................................................'................... Safety Injection SPDS ........................................................................... Safety Parameter Display System SRO ............................................................................................ Senior Reactor Operator SSF ................................................................................................Safe Shutdown Facility TEDE-...... ,...... ,...... ,...... ,...... ,...... ,...... ,...... ,...... ,................ Total Effective Dose Equivalent TOAF ..................................................................................................... Top* of Active Fuel Page 20 of 236 INFORMATION USE I

TSC .......................................................................................... Technical Support Center WOG ................................................................................... Westinghouse Owners Group Page 21 of 236 INFORMATION USE

6.0 CNP-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a CNP EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the CNP EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

CNP NEI 99-01 Rev. 6 Example EAL IC EAL RU1.1 AU1 1, 2 RU1.2 AU1 3 HU2.1 AU2 1 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RG1.1 AG1 1 RG1.2 AG1 2 RG1.3 AG1 3 RG2.1 AG2 1 CU1.1 CU1 1 Page 22 of 236 INFORMATION USE

CNP NEI 99-01 Rev. 6 Example EAL IC EAL CU1.2 CU1 2 CU2.1 . CU2. 1 CU3.1 CU3

• 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1, 2, 3 CA1J CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1, 2 CA6.1 CA6 1 CS1.1 CS1 3 CG1.1 CG1 2 FA1.1 FA1 1 FS1 .1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2 3 HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 Page 23 of 236 INFORMATION USE

CNP NEI 99-01 Rev. 6 Example EAL IC EAL HU4.4 HU4 4 HU7.1 .. HU? 1 HA1.1 HA1 1, 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA? 1

. HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG7.1 HG? 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 2 SU5.1 SU4 1,2,3 SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1,2,3 SU8.1 SU? 1, 2 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA9.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 Page 24 of 236 INFORMATION USE

CNP NEI 99-01 Rev. 6 Example EAL IC EAL SG1.1 SG1 1 SG2.1 . SGS 1 .

EU1.1 E-HU1 1 Page 25 of 236 INFORMATION USE

7.0 ATTACHMENTS 7 .1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Fission Product Barrier Matrix and Basis 7.3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Page 26 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category R - Abnormal Rad Release / Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Many EALs are based on actual or potential degradation of fission product barriers because

• of the elevated potential for offsite radioactivity release. Degradation of fission product

• barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a .failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipme.nt necessary to ensure pla.nt sc1fety:

Events of this category pertain to the following subcategories:

1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable lim.its.

2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.

3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

Page 27 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:

RU1.1 Unusual Event Reading on any Table R-1 effluent radiation monitor> column ".UE" for ~ 60 min.

(Notes 1, 2, 3)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will

_likely be exceeded. _

_ Note 2: If an ol"lgoing release il> dete_cted _and the release__start time is unknown_, assume t_hat the rel_ease _

duration has exceeded the specified time limit.

Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE ALERT UE Unit Vent Noble Gas VRS-1500 (2500) 3.3E+OO µCi/cc 3.3E-01 µCi/cc 3.3E-02 µCi/cc 4.2E-03 µCi/cc (I)

I 0

GI (I)

Gland Seal Leakoff SRA-1800 (2800) 1.6E+02 µCi/cc 1.6E+01 µCi/cc 1.6E+OO µCi/cc 1.4E-01 µCi/cc cu Cl Steam Jet Air Ejector SRA-1900 (2900) 1.5E+04 µCi/cc 1.5E+03 µCi/cc 1.5E+02 µCi/cc 1.3E+01 µCi/cc Radwaste Effluent RRS-1001 ---- -- --- 4.6E+04 cpm R-19 ---- --- ---- 1.7E+03 cpm SG Slowdown

":i er DRS-3100/4100 ---- ---- ---- 1.2E+04 cpm

i R-24 ---- ---- --- 2.9E+04 cpm SG Slowdown Treatment DRS-3200/4200 ---- ---- --- 1.2E+05 cpm Mode Applicability:

All Definition(s):

None Basis:

The column "UE" gaseous and liquid release values in Table R-1 represent two times the appropriate ODCM release rate limits associated with the specified monitors (ref. 1, 2).

This-IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time Page 28 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases (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.

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 *releases, and to control and moilitofintentional 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

. environ*merit is established. If the effluent flow past an effluent monitor is known tcf 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.

This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.

Escalation of the emergency classification level would be via IC RA1.

CNP Basis Reference(s):

1. EP-CALC-CNP-1601, Radiological Effluent EAL Threshold Values

2. PMP-601 O-OSD-001, Off Site Dose Calculation Manual

3. NEI 99-01 AU1 Page 29 of 236 INFORMATION USE

ATTACHMENT 1

\ EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.

EAL:

RU1.2 Unusual Event

.Sample analysis for a gaseous or liquid release indicates a concentration or release rate

> 2 x ODCM limits for;:: 60 min. (Notes 1, -2)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceedeq .

. Note 2: . If an ongojng release is detected anq the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

None Basis:

This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-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.

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 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.

Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.

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 lake water systems, etc.).

Escalation of the emergency classification_ level would be via IC RA1 '.

-1._____________P_a_ge_30_of_2_3_6_ _ _ _ _ _ _ 1N_F_O_R_M_A_T_IO_N_U_S_E__,

ATTACHMENT 1 EAL Bases CNP Basis Reference(s):

1. PMP-601 O-OSD-001, Off Site Dose Calculation Manual

2. NEI 99-01 AU1 Page 31 of 236 INFORMATION USE 1*.

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose

• greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:

RA1.1 Alert Reading on any Table R-1 effluent radiation monitor> column "ALERT" for 2:: 15 min.

(Notes 1, 2, 3, 4)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

• Note 2: . If an ongoing release is. detected and the rel.ease start time is unknown, assume that the release ..

duration has exceeded the specified time limit.

Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Note 4 The pre-calculated effluent monitor values presented in EALs RA 1.1, RS1 .1 and RG 1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE ALERT UE Unit Vent Noble Gas VRS-1500 (2500) 3.3E+OO µCi/cc 3.3E-01 µCi/cc 3.3E-02 µCi/cc 4.2E-03 µCi/cc en

s 0

Q) en Gland Seal Leakoff SRA-1800 (2800) 1.6E+02 µCi/cc 1.6E+01 µCi/cc 1.6E+OO µCi/cc 1.4E-01 µCi/cc

(!)

Steam Jet Air Ejector SRA-1900 (2900) 1.5E+04 µCi/cc 1.5E+03 µCi/cc 1.5E+02 µCi/cc 1.3E+01 µCi/cc Radwaste Effluent RRS-1001 ---- ---- ---- 4.6E+04 cpm R-19 ---- ---- ---- 1.7E+03 cpm

":i SG Slowdown "t:J C" DRS-3100/4100 ---- ---- ---- 1.2E+04 cpm

J R-24 ---- ---- ---- 2.9E+04 cpm SG Slowdown Treatment DRS-3200/4200 ---- ---- ---- 1.2E+05 cpm Mode Applicability:

All Definition(s):

None Page 32 of 236 INFORMATION USE

ATTACHMENT 1 EAL sa*ses Basis:

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:

• 10 mRem TEDE

• .* 50 mRe*m CDEThyroid

• The column "ALERT" gaseous effluent release values in Table R-1 correspond to calculated.

doses of 1% (10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEDE or COE Thyroid) (ref. 1). -

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 !l10nitore9 and un:-monitored re.leases. Releases of this magn_itud~ rep_rese_nt 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 uhcoritrolled release).

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 COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

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 RS1.

CNP Basis Reference(s):

1. EP-CALC-CNP-1601, Radiological Effluent EAL Threshold Values

2. NEI 99-01 AA 1 Page 33 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:

RA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid COE at or beyond the site boundary (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RA 1.1, RS1 .1 and RG 1.1 should be used

. for_ emergency classification assessme11ts until the_ results fro_m a dose ~ssessmen_t using actual meteorology are available.

Mode Applicability:

All Definition(s):

None Basis:

Dose assessments are performed by computer-based or manual methods (ref. 1).

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).

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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE.

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 RS1.

CNP Basis Reference(s):

1. PMP-2080-EPP-108 Initial Dose Assessment 2 .. NEI 99-01 AA1 Page 34 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE

• EAL:**

RA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid COE at or beyond the site boundary for 60 min. of exposure (Notes 1, 2)

Note 1: . The SEC should declare the event promptly upon determining that time limit has been exceeded, or will .

likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

None Basis:

Dose assessments based on liquid releases are performed per Offsite Dose Calculation Manual (ref. 1).

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).

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 COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

Escalation of the emergency classification level would be via IC RS1.

CNP Basis Reference(s):

1. PMP-601 O-OSD-001, Off Site Dose Calculation Manual

. 2. NEI 99-01 AA 1 I Page 35 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose*

greater than 10 mrem TEOE or 50 mrem thyroid COE EAL:

RA1.4 Alert Field survey results indicate EITHER of the following at or beyond the site boundary: .

• Closed window dose rates > 1O mR/hr expected to continue for~ 60 min.

• Analyses of field survey samples indicate thyroid COE> 50 mrem for 60 min. of inhalation.

(Notes 1; 2) **

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

None Basis:

RMT-2080-EOF-001, Activation and Operation of the EOF provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

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).

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 TEOE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEOE and thyroid COE.

Escalation of the emergency classification level would be via IC RS1.

I Page 36 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases CNP Basis Reference(s):

1. RMT-2080-EOF-001 Activation and Operation of the EOF

2. NEI 99-01 AA1 Page 37 of 236 INFORMATION USE . I

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL:

RS1.1 Site Area Emergency Reading on any Table R-1 effluent radiation monitor> column "SAE" .for~ 15 min.

(Notes 1, 2, 3, 4)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit: . . . . . . . . .. . .. . .

Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE ALERT UE Unit Vent Noble Gas VRS-1500 (2500) 3.3E+OO µCi/cc 3.3E-01 µCi/cc 3.3E-02 µCi/cc 4.2E-03 µCi/cc Ul

I 0

Cll Ul Gland Seal Leakoff SRA-1800 (2800) 1.6E+02 µCi/cc 1.6E+01 µCi/cc 1.6E+OO µCi/cc 1.4E-01 µCi/cc ca Cl Steam Jet Air Ejector SRA-1900 (2900) 1.5E+04 µCi/cc 1.5E+03 µCi/cc 1.5E+02 µCi/cc 1.3E+01 µCi/cc Radwaste Effluent RRS-1001 ---- ---- ---- 4.6E+04 cpm R-19 ---- -- ---- 1.7E+03 cpm "C

':i SG Slowdown er DRS-3100/4100 ---- -- ---- 1.2E+04 cpm

i R-24 ---- ---- ---- 2.9E+04 cpm SG Slowdown Treatment DRS-3200/4200 ---- ---- --- 1.2E+05 cpm Mode Applicability:

All Definition(s):

None Page 38 of 236 INFORMATION USE l

ATTACHMENT 1 EAL Bases Basis:

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:

• 100 mRem TEDE

• * *

• 500 mRem COE Thyroid **

The column "SAE" gaseous effluent release value in Table R-1 corresponds to c*alculated doses of 10% of the EPA Protective Action Guidelines (TEDE or COE Thyroid) (ref. 1).

This IC addresses a release of gaseous *radioactivity that results in projected or actual offsite .

. doses greater than or equal to 10% ofthe EPA Protective Action Guides (PAGs). it includes both monitored and un-monitored releases. Releases of this magnitude are associated with th_e fai_lure _of plant syste_ms need_ed for the prqtecti_on o_f the_ pubHc .

.. 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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE.

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 RG1.

CNP Basis Reference(s):

1. EP-CALC-CNP-1601-CNP-1601, Radiological Effluent EAL Threshold Values

2. NEI 99-01 AS1 Page 39 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL:**

RS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > .100 mrem TEDE or 500 mrem thyroid COE at or beyond the site boundary (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RA1 .1, RS1 .1 and RG1 .1 should be used for emergency classification assessmen~s until the resul~s from a d9se assessment using actual meteorology are available.

Mode Applicability:

All Definition(s):

None Basis:

Dose assessments are performed by computer-based and manual methods (ref. 1).

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.

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 COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEDE and thyroid COE.

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 RG1.

CNP Basis Reference(s):

1. PMP-2080-EPP-108 Initial Dose Assessment

2. NEI 99-01 AS1 Page 40 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than*

100 mrem TEOE or 500 mrem thyroid COE

• EAL:

RS1 .3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the site boundary: .

• Closed window dose rates >*100 mR/hr expected to continue for;;::: 60 min.

• Analyses of field survey samples indicate thyroid COE > 500 mrem for 60 min. of inhalation.

(Notes 1; 2)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

None Basis:

RMT-2080-EOF-001, Activation and Operation of the EOF provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

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.

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 TEOE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established in consideration of the 1 :5 ratio of the EPA PAG for TEOE and thyroid COE.

Escalation of the emergency classification level would be via IC RG1.

CNP Basis Reference(s):

1. RMT-2080-EOF-001 ActivaUon and Operatio.n of the EOF.

2. NEI 99-01 AS1 Page 41 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE

• EAL:

RG1.1 General Emergency Reading on any Table R-1 effluent radiation monitor> column "GE" for ~ 15 min ..

(Notes 1, 2, 3, 4)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

• Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.

Note 4: The pre-calculated effluent monitor values presented in EALs RA 1.1, RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE ALERT UE Unit Vent Noble Gas VRS-1500 (2500) 3.3E+OO µCi/cc 3.3E-01 µCi/cc 3.3E-02 µCi/cc 4.2E-03 µCi/cc en

I 0

Q) en Gland Seal Leakoff SRA-1800 (2800) 1.6E+02 µCi/cc 1.6E+01 µCi/cc 1.6E+OO µCi/cc 1.4E-01 µCi/cc Cll C)

Steam Jet Air Ejector SRA-1900 (2900) 1.5E+04 µCi/cc 1.5E+03 µCi/cc 1.5E+02 µCi/cc 1.3E+01 µCi/cc Radwaste Effluent RRS-1001 ---- ---- ---- 4.6E+04 cpm R-19 ---- ---- ---- 1.7E+03 cpm

"::i SG Slowdown C" DRS-3100/4100 ---- ---- ---- 1.2E+04 cpm

i R-24 ---- ---- ---- 2.9E+04 cpm SG Slowdown Treatment DRS-3200/4200 ---- ---- ---- 1.2E+05 cpm Mode Applicability:

All Definition(s ):

None Basis:

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:

Page 42 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases

• 1000 mRem TEDE

• 5000 mRem COE Thyroid The column "GE" gaseous effluent release values in Table R-1 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE or COE Thyroid) (ref. 1).

This IC addresses a rele*ase of gaseous radioactivity that results in projected or actual offsite *

• doses greater than or equal 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.

• 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 PAG of 1;ooo mrem while the 5,000 mreiTI thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

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.

CNP Basis Reference(s):

1. EP-CALC-CNP-1601, Radiological Effluent EAL Threshold Values

2. NEI 99-01 AG1 Page 43 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition:

• Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE

• EAL:

RG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid COE at or beyond the site boundary (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RA 1.1, RS1 .1 and RG1 .1 should be used

. for emergency _class_ification as_sessr:nents until_ the results fro~ a do_se assessment using actual meteorology are available.

Mode Applicability:

All Definition(s):

None Basis:

Dose assessments are performed by computer-based and manual methods (ref. 1).

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 (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

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 PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

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.

CNP Basis Reference(s):

1. PMP-2080-EPP-108 Initial Dose Assessment

2. NEI 99-01 AG1 Page 44 of 236 INFORMATION USE I

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEOE or 5,000 mrem thyroid COE

• EAL:

RG1 .3 General Emergency Field survey results indicate EITHER of the following at or beyond the site boundary:

• Closed window dose rates> 1,000 mR/hr expected to continue for~ 60 min.

• Analyses of field survey samples indicate thyroid COE > 5,000 mrem for 60 min. of inhalation ..

(Notes 1, 2)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Mode Applicability:

All Definition(s):

None Basis:

RMT-2080-EOF-001, Activation and Operation of the EOF provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

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 (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

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 TEOE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEOE and thyroid COE.

CNP Basis Reference(s):

1. RMT-2080-EOF-001 Activation and Operation of the EOF

2. NEI 99-01 AG1 Page 45 of 236 INFORMATION USE

• 1

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event

• Initiating Condition: UNPLANNED loss of water level above irradiated fuel EAL:

RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by low water level alarm or indication AND UNPLANNED rise in corresponding area radiation levels as indicated on any of the following. radi.ation monitors:

.* VRS-1101/1201,Unit1 UpperContainment

• VRS-2101/2201, Unit 2 Upper Containment

• R-5 Spent Fuel Area

• VRS-5006 Spent Fuel Area Mode Applicability:

All Definition(s):

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 comprise the refueling pathway.

Basis:

The low water level alarm in this EAL refers to the Spent Fuel Pool (SFP) low (Panel 134 Drop

2) or low-low level alarms (Panel 105 Drop 27 or Panel 205 Drop 27) (ref. 1). During the fuel transfer phase of refueling operations, the fuel transfer canal is normally in communication with the spent fuel pool and the refueling cavity in the Containment is in communication with the fuel transfer canal when the fuel transfer tube is open. A lowering in water level in the SFP, fuel transfer canal or refueling cavity is therefore sensed by the SFP low level alarm. Neither the refueling cavity nor the fuel transfer canal is equipped with a low level alarm (ref. 1).

Technical Specification Section 3.7.14 (ref. 5) requires at least 23 ft of water above the SFP storage racks. Technical Specification Section 3.9.6 (ref. 4) requires at least 23 ft of water above the Reactor Vessel flange in the refueling cavity. During refueling, this maintains sufficient water level in the fuel transfer canal, refueling cavity, and SFP to retain iodine fission product activity in the water in the event of a fuel handling accident. .

Page 46 of 236 INFORMATION USE ,-

ATTACHMENT 1 EAL Bases The listed radiation monitors are those expected to see increase area radiation levels as a result of a loss of REFUELING PATHWAY inventory (ref. 1, 2, 3). Increasing radiation indications on these monitors in the absence of indications of decreasing REFUELING PATHWAY level are not classifiable under this EAL.

When the spent fuel pool and reactor cavity are connected, there could exist the possibility of uncovering irradiated fuel. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the reactor vessel and spent fuel pool.

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 b.e primari.ly determined by indic~tions from availa.ble level.

instrumentation. Other sources of level indications may include reports from plant personnel (e.g:, fr6m a *refueling crew) of video camera observations (if available).* Asigr\ificailt drop iri the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.

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 RA2.

CNP Basis Reference(s):

1. 12-0HP-0422-018-002 Loss of Refueling Water Level During Refueling Operations - Local Actions

2. 12-0HP-4022-018-003 Irradiated Fuel Handling Accident in Containment Building - Local Actions

3. 12-0HP-4022-018-004 Irradiated Fuel Handling Accident in Containment Building - Control Room Actions

4. Technical Specification Section 3.9.6 Refueling Cavity Water Level

5. Technical Specification Section 3.7.14 Fuel Storage Pool Water Level

6. NEI 99-01 AU2 Page 4 7 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.1 Alert Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:

All

. Definition(~): .

REFUELING PATHWAY -The reactor refueling cavity, spent fuel pool and fuel trans.fer canal comprise the refueling pathway.

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 EU1 .1.

Escalation of the emergency would be based on either Recognition Category R or C ICs.

This EAL escalates from RU2.1 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 (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations.

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.

CNP Basis Reference(s):

1. NEI 99-01 AA2 Page 48 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel

• EAL:

RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by High alarm on any of the following radiation monitors:

• VRS-1101/1201, Unit 1 Upper Containment

• VRS-2101/2201, Unit 2 Upper Containment

• • R-5 Spent Fuel Area

• VRS-5006 Spent Fuel Area ..

Mode Applicability:

All Definition(s):

None Basis:

The specified radiation monitors are those expected to see increase area radiation levels as a result of damage to irradiated fuel (ref. 1, 2, 3, 4).

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 EU1 .1.

Escalation of the emergency would be based on either Recognition Category R or C ICs.

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).

CNP Basis Reference(s):

1. 12-0HP-4022-018-003 Irradiated Fuel Handling Accident in Containment Building - Local Actions Page 49 of 236 INFORMATION USE

.J

ATTACHMENT 1 EAL Bases

2. 12-0HP-4022-018-004 Irradiated Fuel Handling Accident in Containment Building - Control Room Actions

3. 12-0HP-4022-018-005 Irradiated Fuel Handling Accident in Spent Fuel Storage Area - .

Local Actions

4. 12-0HP-4022-018"'.006 Irradiated Fuel Handling Accident in Spent Fuel Storage Area"'.'."

Control Room Actions

5. NEI 99-01 AA2 Page 50 of 236 INFORMATION USE I

1.

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event .

Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.3 Alert Lowering of spent fuel pool level to 9 ft. 6 in. on 1(2)~RLl-502-CRI Spent Fuel Pit Level Indication (8 ft. 10 in. on local ruler)

• Mode Applicability:

All Definition(s):

Nori*e Basis:

Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1).

For CNP SFP Level 2 is plant elevation 630 ft. 10.5 in. or 9 ft. 6 in. as indicated on 1(2)-RLl-502-CRI in the Control Room or 1(2)-RLl-502-BATT back-up indicator (ref. 2). This level corresponds to 8 ft. 10 in. on the SFP ruler.

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.

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 IC RS1 or RS2.

CNP Basis Reference(s):

1. NRG EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. EC-0000052892 Spent Fuel Pool Level for NRG Order EA-12-051 3 NEI 99-01 AA2 Page 51 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL:

RS2.1 Site Area Emergency Lowering of spent fuel pool level to O ft. on 1(2)-RLl-502-CRI Spent Fuel Pit Level Indication Mode Applicability:

All

• Definition(s):

None*

Basis:

Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1).

For CNP SFP Level 3 is plant elevation 620 ft. 10.5 in. However, the SFP level instrument lower range (0 ft.) corresponds to plant elevation 621 ft. 6 in. Therefore an indicated level of 0 ft. on 1(2)-RLl-502-CRI in the Control Room or 1(2)-RLl-502-BATT back-up indicator is used as indicated Level 3 (ref. 2).

This EAL 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 RG1 or RG2.

CNP Basis Reference(s):

1. NRG EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. EC-0000052892 Spent Fuel Pool Level for NRG Order EA-12-051

3. NEI 99-01 AS2 Page 52 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:

RG2.1 General Emergency Spent fuel pool level cannot be restored to at least O ft. on 1(2)-RLl-502-CRI Spent Fuel Pit Level Indication for 2: 60. min. (Note 1)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. -

Mode Applicability:

All Definition(s):

None Basis:

Post-Fukushima order EA-12-051 (ref. 1) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3) (ref. 1).

For CNP SFP Level 3 is plant elevation 620 ft. 10.5 in. However, the SFP level instrument lower range (0 ft.) corresponds to plant elevation 621 ft. 6 in. Therefore an indicated level of 0 ft. on 1(2)-RLl-502-CRI in the Control Room or 1(2)-RLl-502-BATI back-up indicator is used as indicated Level 3 (ref. 2).

This EAL 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.

CNP Basis Reference(s):

1. NRG EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

2. EC-0000052892 Spent Fuel Pool Level for NRG Order EA-12-051

3. NEI 99-01 AG2 Page 53 of 236 INFORMATION USE I . .

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

RA3.1 Alert Dose rates > 15 mR/hr in any of the following areas:

• Unit 1 Control Room (ERS-7401)

• Unit 2 Control Room (ERS-8401)

• Centra_l Ala_rm Station (by survey)

Mode Applicability:

All Definition(s):

None Basis:

Areas that meet this threshold include the Control Rooms and the Central Alarm Station (GAS). ERS-7401 (ERS-8401) monitor the Control Rooms for area radiation (ref. 1). The GAS is included in this EAL because of its importance to permitting access to areas required to assure safe plant operations (ref. 1).

There is no permanently installed GAS area radiation monitors that may be used to assess this EAL threshold. Therefore this threshold must be assessed via local radiation survey for the GAS (ref. 1).

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 SEC should consider the cause of the increased radiation levels and determine if another IC may be applicable.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.

CNP Basis Reference(s):

1. FSAR Table 11.3-1 Radiation Monitoring System Channel Sensitivities and Detecting Medium

2. NEI 99-01 AA3 Page 54 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown EAL:**

RA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any.

Table R-2 rooms or areas (Note 5)

Note 5: 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.

Table R-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Applicability Auxiliary Building 573' BART Area 3,4, 5 Auxiliary Building 587' Boric Acid Storage Tank Room, Nuclear 3,4, 5 Sampling Room 4KV Room (Mezzanine Area), Boric Acid Batch Tank Area, Chemistry 3,4, 5 Hot Lab, RHR Hx Room Mode Applicability:

All Definition(s):

IMPEDE(D) - Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel info the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).

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:

If the equipment in the listed 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 beyond that already allowed by Technical Specifications at the time of the event.

The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. 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) are not included. In addition, the list specifies the plant

. mode(s) during which entry would be required for each room or area (ref. 1).

Page 55 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 SEC should consider the cause of the increased radiation levels and determine if another IC may be applicable. For this EAL, 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 .

roo.m/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.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.

CNP Basis Reference(s):

1. Attachment 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases

2. NEI 99-01 AA3 Page 56 of 236 INFORMATION USE -1

ATTACt-JMENT 1 EAL Bases Category E - Independent Spent Fuel Storage Installation (ISFSI)

EAL Group: Any (EALs in this category are applicable to any plant condition, hot or cold.)

An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained

.within a canister must escape its packaging and enter the biosphere for there to be a . .

significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel.

An UNUSUAL EVENT is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated.

Page 57 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: ISFSI Subcategory: Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:

I EU1.1

• Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading:

• > 60 mrem/hr (gamma + neutron) on the top of the overpack

• > 600 mrem/hr (gamma + neutron) on the side of the overpack excluding inlet and

• outlet ducts*

Mode Applicability: I All Definition(s):

CONFINEMENT BOUNDARY-. The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As applied to the CNP ISFSI, the CONFINEMENT BOUNDARY is defined to be the Multi-Purpose Canister (MPC).

Basis:

Overpacks are the HI-STORM 100 casks which receive and contain the sealed MPCs for interim storage in the ISFSI. They provide gamma and neutron shielding, and provide for ventilated air flow to promote heat transfer from the MPC to the environs. The term overpack does not include the transfer cask (ref. 1).

The value shown represents 2 times the maximum overpack surface dose rates specified in Section 5.7 of the ISFSI Certificate of Compliance Technical Specifications for radiation external to a loaded MPC overpack (ref. 1).

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 RU1, 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.

Page 58 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Security-related events for ISFSls are covered under ICs HU1 and HA1.

CNP Basis Reference(s):

1. Certificate of Compliance No. 1014 Holtec International HI-STORM 100 Cask System Safety Evaluation Report Amendment 1 Appendix A Technical Specifications Section 5.7

. 2. NEI 99.,01 E-HU1 Page 59 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category C - Cold Shutdown I Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature s 200°F); EALs in this category are applicable only in one or more cold operating modes .

. Category C EALs are. directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown)during these periods, the consequences of any given initiating event can vary greatly. For example, a

• loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (5 -

Cold Shutdown, 6 - Refueling, D - Defueled).

The events of this category pertain to the following subcategories:

1. RCS Level RCS water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.

2. Loss of Emergency AC Power Loss of emergency plant electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4.16KV AC emergency buses.

3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.

4. Loss of Vital DC Power Loss of emergency electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of vital plant 250 VDC power sources.

5. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in VISIBLE DAMAGE to or degraded performance of SAFETY SYSTEMS warranting classification.

Page 60 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory EAL:

CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RCS water level less than a required lower limit for.;:: 15. min. (Note 1)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

• 5 - Cold Shutdown, 6 - Refueling ..

Definition(s):

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:

With the plant in Cold Shutdown, RCS water level is normally established by 1 (2)-0HP-4021-002-005, RCS Draining (ref. 1). If RCS level is being controlled below the pressurizer low level setpoint, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern.

With the plant in Refueling mode, RCS water level is normally maintained at or above the reactor vessel flange (Technical Specification LCO 3.9.6 requires at least 23 ft. of water above the top of the reactor vessel flange in the refueling cavity during refueling operations) (ref. 2).

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 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.

This EAL recognizes that the minimum required 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 m*inimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.

• 1~_ _ _ _ _ _ _ _ _ _ _ _P_a_g_e_6_1_of_2_3_6_ _ _ _ _ _ _ 1N_F_O_R_M_A_T_IO_N_U_S_E~

ATTACHMENT 1 EAL Bases 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.

Continued loss of RCS inventory may result in escalation to the ALERT emergency classification level via either IC CA1 or CA3.

CNP Basis Reference{s):

1. 1(2)-0HP-4021-002-005, RCS Draining

2. Technical Specification Section 3.9.6 Refueling Cavity Water Level

3. NEI 99-01 CU1 Page 62 of 236 INFORMATION USE . ,

. I

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: UNPLANNED loss of RCS inventory EAL:

CU1.2 Unusual Event RCS water level cannot be monitored

- AND EITHER

• UNPLANNED increase in any Table C-1 sump/tank level due to loss of RCS inventory

• Visual observation of UNISOLABLE RCS leakage Table C-1 Sumps / Tanks

• Containment Sumps

• Auxiliary Building Sumps

• RWST

• RCDT Mode Applicability:

5 - Cold Shutdown, 6- Refueling Definition(s):

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.

Basis:

In Cold Shutdown mode, the RCS will normally be intact and standard RCS level monitoring means are available.

In this EAL, all water level indication is unavailable and the RCS inventory loss must be detected by indirect leakage indications. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2).

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 RCS level concurrent with Page 63 of 236 INFORMATION USE--,

ATTACHMENT 1 EAL Bases 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.

This EAL addresses a* condition where all means to determine 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 (Table C-1 ). 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 RCS.

Conti_nued loss of RCS_ inventory may result in es~alation to the ALERT_ em~rgency classification level via either IC CA1 or CA3.

CNP Basis Reference(s):

1. 1(2)-0HP-4022-002-020 Excessive Reactor Coolant Leakage

2. 1(2)-0HP-4021-002-005, RCS Draining

3. NEI 99-01 CU1 Page 64 of 236 INFORMATION USE I

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Significant loss of RCS inventory*

EAL:

CA1.1 Alert Loss of RCS inventory as indicated by RCS level< 614.0 ft .

. Mode Applicability: .

5 - Cold Shutdown, 6- Refueling Definition(s):

None.

Basis:

614.0 ft. corresponds to midloop and is the minimum allowed RCS level for operation of RHR (ref.1)

RCS level cannot be measured below 612 feet on NLl-1000, RCS Half-Loop Operation Wide Range Level Indication, which is below the bottom ID of the hot leg inlet. Should RCS level drop below this point it is assumed water level cannot be monitored.

This IC addresses conditions that are precursors 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 this EAL, a lowering of RCS water level below 614.0 ft. indicates that operator actions have not been successful in restoring and maintaining 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.

Although related, this EAL 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.

If RCS water level continues to lower, then escalation to SITE AREA EMERGENCY would be via IC CS1.

CNP Basis Reference(s):

1. 1(2)-0HP-4022-017-001 Loss of RHR Cooling

2. NEI 99-01 CA1 Page 65 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition:

• Significant loss of RCS inventory EAL:

CA1.2 Alert RCS water level cannot be monitored for;;;: 15 min. (Note 1)

AND EITHER

• UNPLANNED increase in any Table C-1 sump/tank level due to loss of RCS inventory

• Visual observation of UNISOLABLE RCS leakage Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table C-1 Sumps / Tanks

• Containment Sumps

• Auxiliary Building Sumps

• RWST

• RCDT Mode Applicability:

5 - Cold Shutdown, 6- Refueling Definition(s):

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.

Basis:

In Cold Shutdown mode, the RCS will normally be INTACT and standard RCS level monitoring means are available.

In the Refuel mode, the RCS is not INTACT and RPV level may be monitored by different means, including the ability to monitor level visually.

In this EAL, all RCS water level indication would be unavailable for greater than 15 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-1).

Surveillance procedures provide instructions for calculating primary system leak rate by manual or computer-based water inventory balances. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS Page 66 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2).

This IC addresses conditions that are precursors to a loss of the ability to adequately cool

• irradiated fu*e1 (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 this EAL, the inability to monitor 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 of leakage from the RCS.

The 15-minute duration for the loss of level indication was chosen because it is half of the* EAL duration specified in IC CS1.

If the RCS inventory level continues to lower, then escalation to SITE AREA EMERGENCY would be via IC CS1.

CNP Basis Reference(s):

1. 1(2)-0HP-4022-002-020 Excessive Reactor Coolant Leakage

2. 1(2)-0HP-4021-002-005, RCS Draining

3. NEI 99-01 CA 1 Page 67 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS1 .1 Site Area Emergency RCS water level cannot be monitored for~ 30 min. (Note 1) .

AND Core uncovery is indicated by any of the following:

• UNPLANNED increase*in any Table*C-1 sump/tank level of sufficient magnitude to

• indicate core uncovery

• High alarm on Containment radiation monitorVRA-1310 (2310) orVRA-1410(2410)

• Erratic Source Range Monitor indication Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table C-1 Sumps / Tanks

• Containment Sumps

• Auxiliary Building Sumps

• RWST

• RCDT Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Definition(s):

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.

Basis:

In Cold Shutdown mode, the RCS will normally be INTACT and standard RCS level monitoring means are available.

In the Refueling mode, the RCS is not INTACT and reactor vessel level may be monitored by different means, including the ability to monitor level visually.

In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-1 ).

Surveillance procedures provide instructions for calculating primary system leak rate by Page 68 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases manual or computer-based water inventory balances. Level increases must be evaluated against other potential sources of leakage such as cooling water s0urces inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre.:.established rate, *a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could also be indicative of a los~ of RCS inventory (ref. 1, _2).

The reactor vessel inventory loss may be detected by the containment radiation monitors VRA-1310 (2310) or 1410 (2410) or erratic Source Range Monitor indication: As water level in the re~ctor vessel lowers, the dose rate above the core will rise. The dos.e rate due to .this core shine should result in a high alarm on containment high range radiation monitors (ref. 3).

Post-TM I 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 (ref. 4).

This IC addresses a significant and prolonged loss of 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 level cannot be restored, fuel damage is probable.

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.

The inability to monitor 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 of leakage from the RCS.

This EAL addresses 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 CG1 or RG1 Page 69 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases CNP Basis Reference(s):

1. 1 (2)-0HP-4022-002-020 Excessive Reactor Coolant Leakage

2. 1(2)-0HP-4021-002-005, RCS Draining

3. ECP 1-2-V2-01 Post Accident High Range Containment Area Radiation Monitoring

4. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island - Unit 2 Accident," NSAC-1 *

5. NEI 99-01 CS1 Page 70 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RCS Level Initiating Condition: Loss of RCS inventory affecting fuel clad integrity with containment challenged EAL:

CG1 .1 General Emergency RCS level cannot be monitored for;.:: 30 min. (Note 1)

AND Core uncovery is indicated by any of the following:

• UNPLANNED increase in *any Table c..:1 sump/tank level of sufficient magnitude to*

indicate core uncovery *

• High alarm on Containment radiation monitorVRA-1310 (2310) orVRA-1410(2410)

• Erratic Source Range Monitor indication AND Any Containment Challenge indication, Table C-2 Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a GENERAL EMERGENCY is not required.

Table C-1 Sumps / Tanks

• Containment Sumps

• Auxiliary Building Sumps

• RWST

• RCDT Table C-2 Containment Challenge Indications

• CONTAINMENT CLOSURE not established (Note 6)

• Containment hydrogen concentration;.:: 4%

• Unplanned rise in Containment pressure Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Page 71 of 236 INFORMATION USE ,.

ATTACHMENT 1 EAL Bases Definition{s):

CONTAINMENT CLOSURE - The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions .

. As applied to CNP, Containment Closure is established when the requirements of PMP:-4100-SDR-001 are met.

UN/SOLABLE - 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.

Basis:

In. Colc;i Shµtdown mode., the. RCS will normally be INTACT.and.stan.darcl RCS level monitoring means are available.

In the Refueling mode, the RCS is not INTACT and RPV level may be monitored by different means, including the ability to monitor level visually.

In this EAL, all RCS water level indication would be unavailable for greater than 30 minutes, and the RCS inventory loss must be detected by indirect leakage indications (Table C-1 ).

Surveillance procedures provide instructions for calculating primary system leak rate by manual or computer-based water inventory balances. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. If the make-up rate to the RCS unexplainably rises above the pre-established rate, a loss of RCS inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS that cannot be isolated could also be indicative of a loss of RCS inventory (ref. 1, 2).

The reactor vessel inventory loss may be detected by the containment radiation monitors VRA-1310 (2310) or 1410 (2410) or erratic Source Range Monitor indication. As water level in the reactor vessel lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in a high alarm on containment high range radiation monitors (ref. 3).

Post-TMI 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 (ref. 4).

Three conditions are associated with a challenge to containment integrity:

1. CONTAINMENT COSURE not established - The status of CONTAINMENT CLOSURE is tracked if plant conditions change that could raise the risk of a fission product release as a result of a loss of decay heat removal (ref. 5). If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute core uncovery time limit then escalation to GE would not occur.

2. Containment hydrogen ~ 4% - The 4% hydrogen concentration threshold is generally considered the lower limit for hydrogen deflagrations. CNP is equipped with a Post-Accident Containment Hydrogen Monitoring System (PACHIVIS) that is capable of continuously measuring the concentration of hydrogen in the containment atmosphere Page 72 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases following a significant beyond design-basis accident for accident mitigation, including emergency planning. PACHMS is comprised of two sampling-analyzing-control trains.

Each train has two subsystems - the hydrogen analyzer panels and the remote control panels (ref. 6).

3. UNPLANNED rise in Containment pressure - An UNPLANNED pressure rise in containment while in cold Shutdown or Refueling modes can threaten CONTAINMENT.

CLOSURE capability and thus containment potentially cannot be relied upon as a barrier to fission product release. * *

• This IC addresses. the inability to restore. and maintain reactor vessel level above the top of active fµel 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 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 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 re-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.

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.

The inability to monitor 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 _cha_nges 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 RCS.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; I Page 73 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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.

CNP Basis Reference(s):

. 1. 1(2),-0HP-4022"'.002"'.020 Excessive Reactor Coolant Leakage

2. 1(2)-0HP-4021-002.,005, RCS Draining

3. ECP 1-2-V2-01 Post Accident High Range Containment Area Radiation Monitoring

4. Nuclear Safety Analysis Center (NSAC), 1980, "Analysis of Three Mile Island :. Unit 2 Accident," NSAC-1 . . . . . .

5. PMP-41 OO-SDR-001 Plant Shutdown Safety and Risk Management

6.

• UFSAR Section 7.8.2 Post-Accident Containment Hydrogen Monitorin*g

7. NEI 99-01 CG1 Page 74 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer

• EAL:

CU2.1 Unusual Event AC power capability, Table C-3, t_o emergency 4.16 kV buses T11A {T21A) and T11 D

{T21D) reduced to a single power source for~ 15 min. (Note 1)

AND Any additiorial s_ingle power source fai_lure will result in loss of au AC power to SAFETY SYSTEMS Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table C-3 AC Power Sources Offsite:

• Reserve Auxiliary Xmr TR101AB (TR201AB)

• Reserve Auxiliary Xmr TR101 CD (TR201 CD)

• 69/4.16 kV Alternate Xmr TR12EP-1

• Main Xmr TR1 (TR2) backfeed (only if already aligned)

Onsite:

• EOG 1AB (2AB)

• EOG 1CD {2CD)

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, D - Defueled Definition(s):

SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result

. in potential *offsite exposures. . . . . . . . . . . . .

Page 75 of 236 INFORMATION USE I.

ATTACHMENT 1 EAL Bases Basis:

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.

The condition indicated by this EAL is the degradation of the offsite and onsite power sources such that any additional singlefailure would result in a loss of all AC power to the emergency buses.

A list of onsite arid offsite AC power sources credited for this EAL are specified in Table C-3.

4.16KV buses T1.1A (T21A) and T11 D (T21 D) are the emergency (essential) buses (ref. 1).

While generating, auxiliary power is normally supplied from the generator terminals through the unit auxiliary transformers (TR1AB and TR1 CD for Unit 1 and TR2AB and TR2CD for Unit 2).

When the plant trips or the plant is shutdown the station auxiliaries are transferred to the preferred*offsite power source (that is, to reserve auxiliary transformers TR101AB and TR 101 CD for Unit 1 and TR201 AB and TR201 CD for Unit 2) to assure continued power to .

equipment when the main generator is off-line (ref. 1).

In addition, an alternate offsite power source, a 69/4.16kV transformer {TR 12EP-1 ), located at the plant site, has the necessary capacity to operate one train of the engineered safeguard equipment in one unit while supplying one train of the safe shutdown power in the other.

T11A {T21A) and T11 D {T21 D) also each have an emergency diesel generator which supply onsite electrical power to the bus automatically in the event that the preferred offsite sources become unavailable (ref. 1).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer and unit auxiliary transformers. This is only done during cold shutdown when no other power sources are available (ref. 1, 3). Credit is only taken for this source if already aligned as it requires removal of the main generator disconnect links.

The Supplemental Diesel Generators (SDGs) are not credited as an AC power source for this EAL.

This cold condition EAL is equivalent to the hot condition EAL SA 1.1.

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.

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 degradation of the level of safety of the plant. .

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an essential bus. Some examples of this 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 Page 76 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 ormomentary losses of ..

power.

The subsequent loss of the remaining single power source would escalate the event to an ALERT in accordance with IC CA2.

• CNP Basis Reference(s):

1. UFSAR Figure 8.1-1A(B) Main Auxiliary One-Line Diagram

2. UFSAR Section 8.0 Electrical Systems

3. 1(2)-0HP-4022-001-005 Loss of Offsite Power with Reactor Shutdown*

4. NEI 99-01 CU2 Page 77 of 236 INFORMATION USE

. I

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL:

CA2.1 Alert Loss of all offsite and all on site AC power to emergency 4.16KV buses T11 A (T21 A) and .

T11 D (T21 D) for~ 15 min. (Note 1)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will

. likely be ~xceeded ..

Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, D - Defueled Basis:

4.16KV buses T11A (T21A) and T11 D (T21 D) are the emergency (essential) buses (ref. 1).

While generating, auxiliary power is normally supplied from the generator terminals through the unit auxiliary transformers (TR1AB and TR1 CD for Unit 1 and TR2AB and TR2CD for Unit 2).

When the plant trips or the plant is shutdown the station auxiliaries are transferred to the preferred offsite power source (that is, to reserve auxiliary transformers TR101AB and TR101CD for Unit 1 and TR201AB and TR201CD for Unit 2) to assure continued power to equipment when the main generator is off-line (ref. 1).

In addition, an alternate offsite power source, a 69/4.16kV transformer (TR12EP-1), located at the plant site, has the necessary capacity to operate one train of the engineered safeguard equipment in one unit while supplying one train of the safe shutdown power in the other.

T11A (T21A) and T11 D (T21 D) also each have an emergency diesel generator which supply electrical power to the bus automatically in the event that the preferred offsite sources become unavailable (ref. 1).

Another method to obtain offsite power is by backfeeding the emergency buses through the main transformer and unit auxiliary transformers. This is only done during cold shutdown when no other power sources are available (ref. 1, 3). Credit is only taken for this source if already aligned as it requires removal of the main generator disconnect links.

The Supplemental Diesel Generators (SDGs) or any other alternative AC power source capable of powering an emergency bus can also be credited as an AC power source for this EAL.

This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SS1.1.

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.

Page 78 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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.

• Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via IC CS1 or RS1.

• CNP Basis Reference(s):

1. UFSAR Figure 8.1-1A(B) Main Auxiliary One-Line Diagram

2. UFSAR s.ection 8.0 Electrical Systems 3 .. 1(2)-0HP-4022-001 "'.005 Loss of Offsite Power with Reactor Shutdown

4. NEI 99-01 CA2 Page 79 of 236 INFORMATION USE

_J

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.1 Unusual Event UNPLANNED increase in RCS temperature to > 200°F (Note 10)

Note 10: Begin monitoring hot condition EALs concurrentiy for *any riew event or condition not related to the loss

• of decay heat removal. * * * * * * * * *

• Mode Applicability:

5 - Cold Shutdown, 6 -"Refueling Definition(s):

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:

The following instrumentation is capable of providing indication of an RCS temperature rise that approaches the Technical Specification cold shutdown temperature limit of (200° F) (ref. 1, 2):

• NTl-100, NTl-101, Selected lncore Temperature or Temporary Thermocouples

• NTR-210, Reactor Coolant T-Cold Wide Range Loop 1

• NTR-220, Reactor Coolant T-Cold Wide Range Loop 2

• NTR-230, Reactor Coolant T-Cold Wide Range Loop 3

• NTR-240, Reactor Coolant T-Cold Wide Range Loop 4

• NTR-110, Reactor Coolant T-Hot Loop 1

• NTR-120, Reactor Coolant T-Hot Loop 2

• NTR-130, Reactor Coolant T-Hot Loop 3

• NTR-140, Reactor Coolant T-Hot Loop 4

• RHR display on PPG In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on time to boil data. Refer to OHP-4022-017-001, Loss of RHR Cooling (ref.2).

This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limitand 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 SEC 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.

This EAL 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 I Page 80 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss offorced decay heatremoval at REDUCED INVENTORY may-result in a rapid increase in reactor coolant temperature depending on the .

time after shutdown.

Escalation to ALERT would be via IC CA 1 based on an inventory loss or IC CA3 based on

• exceeding plant configuration-specific time criteria.

• CNP Basis Reference(s):

1; 1(2)-0H P-4021-001-004, Plant Coo Id own *from Hot Standby to Cold Shutdown

• 2. 1(2).:QHP-4022-017-001, loss of RHR Cooling

3. NEI 99-01 CU3 Page 81 of 236 INFORMATION USE . I

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.2 Unusual Event Loss of all RCS temperature and RCS level indication for~ 15 min. (Note 1)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or wili

• likely be exceeded. * * * * * * * * * * * * * *

• Mode Applicability:

5 - Cold Shutdown, 6- Refueling Definition(s):

None Basis:

The following instrumentation is capable of providing indication of an RCS temperature rise that approaches the Technical Specification cold shutdown temperature limit of (200° F) (ref. 1, 2):

• NTl-100, NTl-101, Selected In core Temperature or Temporary Thermocouples

• NTR-210, Reactor Coolant T-Cold Wide Range Loop 1

• NTR-220, Reactor Coolant T-Cold Wide Range Loop 2

• NTR-230, Reactor Coolant T-Cold Wide Range Loop 3

• NTR~240, Reactor Coolant T-Cold Wide Range Loop 4 o NTR-110, Reactor Coolant T-Hot Loop 1

• NTR-120, Reactor Coolant T-Hot Loop 2

• NTR-130, Reactor Coolant T-Hot Loop 3

• NTR-140, Reactor Coolant T-Hot Loop 4

• RHR display on PPG RCS level indications include pressurizer level, narrow and wide range RVLIS and RC Loop narrow, mid and wide range instruments, NGG-100 and Mansell level instrument (ref. 2.3.4).

This EAL addresses the inability to determine RCS temperature and level, and 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 SEC should also refer to ICCA3.

This EAL 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 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.

Page 82 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to ALERT would be via IC CA 1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

CNP Basis Reference(s):

1. _1(2)-0HP-4021-001-004, Plant Cooldown from Hot Standby to Cold Shutdown

2. 1(2)-0HP-4022-017-001, Loss of RHR Cooling

3. 1(2)-0HP-4022-002-020 Excessive Reactor Coolant Leakage

4. 1(2)-0HP-4021-002-005, RCS Draining

5. NEI 99-01 CU3 Page 83 of 236 INFORMATION USE

__ ___J

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL:

CA3.1 Alert UNPLANNED increase in RCS temperature to> 200°F for> Table C-4 duration (Notes 1, 10)

OR UNPLANNED RCS pressure increase > 10 psig (This EAL does not apply during water-solid plant condit_ions_

Note 1:

• The SEC should declare the event promptly upon determining that the applicable time has been*

exceeded, or will likely be exceeded.

Note 1O: Begin monitoring hot condition EALs concurrently for any new event or condition not related to the loss of decay heat removal.

Table C-4: RCS Heat-up Duration Thresholds CONTAINMENT RCS Status Heat-up Duration CLOSURE Status INTACT (but not N/A 60 min.*

REDUCED INVENTORY)

Not INTACT established 20 min.*

OR REDUCED INVENTORY not established O min.

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

Mode Applicability:

5 - Cold Shutdown, 6- Refueling Definition(s):

CONTAINMENT CLOSURE - The procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions.

As applied to CNP, Containment Closure is established when the requirements of PMP-41 OO-SDR-001 are met.

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.

Page 84 of 236 INFORMATION USE

• 1

ATIACHMENT 1 EAL Bases INTACT (RCS) - The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).

REDUCED INVENTORY - Operating condition when fuel is in the reactor vessel and Reactor Coolant System level is lower than 3 feet (or more) below the Reactor Vessel flange.

Basis:

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (200°F, ref. 1). These include (ref. 2, a): . . . . . . . . . . . . . . .

• NTl-100, NTl-101, Selected In core Temperature or Temporary Thermocouples

• NTR-210, Reactor Coolant T-Cold Wide Range Loop 1

• NTR-220,* Reactor Coolant T-Cold Wide Range Loop 2

• • NTR-230, Reactor Coolant T-Cold Wide Range Loop 3 *

• NTR-240, Reactor Coolant T-Cold Wide Range Loop 4

• NTR-110, Reactor Coolant T-Hot Loop 1

• NTR-120, Reactor Coolant T-Hot Loop 2

• NTR-130, Reactor Coolant T-Hot Loop 3

• NTR-140, Reactor Coolant T-Hot Loop 4

• RHR display on PPG The following instrumentation is capable of providing indication of a 10 psig rise in RCS pressure:

• NLl-1 OOOA/B, RCS Pressure

• NLl-122A/B (MLMS Cart C), RCS Pressure

• NPS-110 (Loop 1) Reactor Vessel Train A Wide Range Pressure

• NPS-111 (Loop 3) Reactor Vessel Train B Wide Range Pressure In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RCS pressure increase criteria when in Mode 5 or based on time to boil data when in Mode 6. Refer to OHP-4022-017-001, Loss of RHR Cooling (ref. 3).

This IC addresses conditions involving a loss 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.

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). 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 Page 85 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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, arid

2) there is reduced reactor coolant inventory above the top .of irradiated fuel. .

The RCS pressure increase threshold provides a pressure-based indication of RCS heat-up.

Escalation of ttie emergency classification level would be via IC CS1 or RS1.

CNP Basis Reference(s):

1.. CNP Technical Specifications Table 1.1-1

2. 1(2)-0HP-:4021-001-004, Plant Cooldown froi:n HQt Standl:>y to .Colq ShLJtdown

3. 1(2)-0HP-4022-017-001, Loss of RHR Cooling

4. NEI 99-01 CA3 Page 86 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer EAL:

CU4.1 Unusual Event

< 215 VDC bus voltage indications on Technical Specification required 250 VDC vital buses for~ 15 min. (Note 1)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicabil.ity: .

5 .- Cold Shutdown, 6 - Refueling Definition(s):

None Basis:

The purpose of this EAL is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. The fifteen minute interval is intended to exclude transient or momentary power losses.

The vital DC buses are the following 250 VDC Class 1E buses (ref. 2, 3):

Train A: Train B:

1CD (2CD) 1AB (2AB)

There are two, 116 cell, lead-acid storage batteries (1AB (2AB) and 1CD (2CD)) that supplement the output of the battery chargers. They supply DC power to the distribution buses when AC power to the chargers is lost or when transient loads exceed the capacity of the battery chargers (ref. 3).

CNP Technical Specification LCO 3.8.5 requires that one Train A or Train B 250 VDC electrical power subsystem shall be OPERABLE to support one train of the DC Electrical Power Distribution System required by LCO 3.8.10, "Distribution Systems - Shutdown." (ref. 1).

Per SD-DCC-NEEP-104, a 210 VDC lower limit has been identified from the battery service test acceptance criteria. Based on interpolation, the low voltage limit that would provide a 15 minute margin has been determined to be 213 VDC (ref. 4).

An EAL value of 215 VDC has been selected to account for available instrument accuracy.

Meter scaling on installed control room instrumentation (10 VDC divisions on a dial indicator) limits the closest value that can be accurately read on the control board to 5 VDC.

This EAL is the cold condition e_quivalent of the hot condiUon loss of DC power EAL SS7.1.

Page 87 of 236 INFORMATION USE

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ATTACHMENT 1 EAL Bases 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 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.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.*

. Depending upori the evelit, escalation of the emergenci classification level would be via IC.

CA 1 or CA3, or an IC in Recognition Category R.

CNP Basis Reference(s):

1. Technical Specifications Section 3.8.5 DC Sources - Shutdown

2. UFSAR Figure 8.3-2

3. UFSAR Section 8.3.4 250 Volt DC System (Safety Related)

4. SD-DCC-NEEP-104 250 voe System

5. NEI 99-01 CU4 Page 88 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown I Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

CUS.1 Unusual Event Loss of all Table C-5 onsite communication methods OR Loss of all Table C-5 ORO communication methods OR Loss of all Table C-5 NRC communication methods Table C-5 Communication Methods System Onsite ORO NRC Plant Page X Plant Radios X X Plant Telephone X X X ENS Line X X Commercial Telephone X X Microwave Transmission X X Mode Applicability:

5 - Cold Shutdown, 6 - Refueling, D- Defueled Definition{s):

None Basis:

Onsite/offsite communications include one or more of the systems listed in Table C-5 (ref. 1).

This EAL is the cold condition equivalent of the hot condition EAL SU7.1.

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 OROs and the NRC. -

Page 89 of 236 INFORMATION USE

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ATTACHMENT 1 EAL Bases 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.).

The first EAL condition addresses a total loss of the communications methods used in support

• of routine plant operations .

. The second EAL condition addres*ses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State and Berrien County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

CNP Basis Reference(s):

1. CNP Plant Emergency Plan Section F Emergency Communications

2. NEI 99-01 CU5 Page 90 of 236 INFORMATION USE I. .J

ATTACHMENT 1 EAL Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:

CA6.1 Alert The occurrence of any Table C-6 hazardous event _

AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operat_ing mode AND EITHER: .

• Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode

• Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 11, 12)

Note 11: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 12: 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.

Table C-6 Hazardous Events

• Seismic event (earthquake)

• Internal or external FLOODING event

• High winds or tornado strike

• FIRE

• EXPLOSION

• Other events with similar hazard characteristics as determined by the SEC Mode Applicability:

5 - Cold Shutdown, 6 - Refueling Page 91 of 236 INFORMATION USE J

ATTACHMENT 1 EAL Bases Definition(s):

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 require a post-event inspection to determine if the attributes of an explosi6n a:re 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 if large quantities of smoke and heat are observed.

FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or 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 are typically systems classified as safety-related (as defined in 10CFRS0.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

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.

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; 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 traln*. . . . . . . . . .

• . _ I_ _ _ _ _ _ _ _ _ _ _ _ P_a_ge_92_of_2_3_6_ _ _ _ _ _ _ 1N_F_O_R_M_A_T_IO_N_U_S_E~

ATTACHMENT 1 EAL Bases 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 CS1 or RS1 ..

CNP Basis Reference(s):

1. EP FAQ 2016-002

2. NEI 99-01 CA6 Page 93 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Hazards are non-plant, system-related events that can directly or indirectly affect plant

. operation, reactor plant safety or personnel safety ..

1. Security Unauthorized entry attempts into the PROTECTED AREA, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.

2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.

3.

• Natural or Technology Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.

4. Fire Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the site PROTECTED AREA or which may affect operability of equipment needed for safe shutdown

5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.

6. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for mqnitoring and controlling plant functions is necessary through the emergency response facilities.

7. SEC Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the SEC the latitude to classify emergency conditions consistent with the established classification criteria based upon SEC judgment.

Page 94 of 236 INFORMATION USE j

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Supervisor.

OR Notification of a credible security threat directed at the site OR . . . . . . .

A validated notification from the NRG providing information of an aircraft threat Mode Applicability:

All Definition(s):

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 CNP 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 CNP. 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:

The security shift supervision is defined as the Security Shift Supervisor.

This EAL is based on the Donald C. Cook Nuclear Plant Security Plan (ref. 1).

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 HA 1, and HS1.

Timely and accurate communications between Security Shift Supervision 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 I Page 95 of 236 INFORMATION USE I .

ATTACHMENT 1 EAL Bases for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan.

The first threshold references the Shift Security Supervisor because these are the individuals trained to confirm that a security event is occurring or has occurred (ref. 1). Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.

The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed iri accordance with the CNP Plant Security Plan and DBT.

The third threshold addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communfcate 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 Donald C. Cook Nuclear Plant S_ecurity P_lan ..

. 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 HA 1.

CNP Basis Reference(s):

1. Donald C. Cook Nuclear Plant Security Plan (Safeguards)

2. NEI 99-01 HU1 Page 96 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:**

HA1.1

• Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor.

OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site

. Mode Applicability: ..

All Definition(s):

HOSTILE ACTION - An act toward CNP 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 CNP. 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 - Area outside the PROTECTED AREA fence that immediately surrounds the plant. Access to this area is generally restricted to those entering on official business.

Basis:

The security shift supervision is defined as the Security Shift Supervisor (ref. 1).

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 the 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) .. The ALERT declaration will also heighten the awareness of Offsite Response Page 97 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Organizations (OROs), 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 bf events is adequately addressed by other EALs, ot the***

requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

The first threshold 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.

The second threshold addresses the threat from the impact of an aircraft on the plant, and the

. anUcipated arrival time is within 30 minutes. The intent 9f thi.s EAL is to ensure tha_t thr~at-.

related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This* EAL is met when*thethreat-related information has been*

validated in accordance with site-specific security procedures.

The NRG 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 NRG.

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 NRG. 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.

CNP Basis Reference(s):

1. Donald C. Cook Nuclear Plant Security Plan (Safeguards)

2. NEI 99-01 HA1

• 1~___________ P_a_ge_98_of_2_3_6_ _ _ _ _ _ _ 1N_F_O_R_M_A_T_IO_N_U_S_E~

ATTACHMENT 1 EAL Bases Category: H - Hazards Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the plant PROTECTED AREA EAL:

HS1 .1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the plant PROTECTED AREA as reported by the Security Shift Supervisor Mode Applicability:

All Definition(s):

HOSTILE ACTION -An act toward CNP 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 CNP. 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).

PROTECTED AREA - The area encompassed by physical barriers to control access to the plant and to the ISFSI.

Basis:

The security shift supervision is defined as the Security Shift Supervisor (ref. 1).

These individuals are the designated onsite personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Donald C.

Cook Nuclear Plant Security Plan (Safeguards) information.

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 Supervision and the Control Room is essential for proper classification of a security-related event (ref. 2, 3, 4 5).

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). The SITE AREA EMERGENCY declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public I Page 99 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 HA 1. It also does not apply to incidents that are accidental events, acts of civil disobedience, a

or otherwise are not a: HOSTILE ACTION perpetrated by HOSTILE FORCE. *Examples include the crash of a small aircraft, shots f_rom 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 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 FG1.

CNP Basis Reference(s):

1. Donald C. Cook Nuclear Plant Security Plan (Safeguards)

2. NEI 99-01 HS1 Page 100 of 236 INFORMATION USE l,.

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than QBE level EAL:

HU2.1 . Unusual Event .

Control Room personnel feel an actual or potential seismic event AND The occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager

. Mode Ptpplicability:_

All Definition(s):

None Basis:

Event verification with external sources should not be necessary during or following an QBE.

Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.1 Og). The Shift Manager or Site Emergency Coordinator 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 (ref. 2).

Ground motion acceleration of 0.1 Og horizontal is the Operating Basis Earthquake for CNP (ref. 1).

To avoid inappropriate emergency classification resulting from spurious actuation of the seismic instrumentation or felt motion not attributable to seismic activity, an offsite agency (USGS, National Earthquake Information Center) can confirm that an earthquake has occurred in the area of the plant. The NEIC can be contacted by calling (303) 273-8500. Select option

1. 1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of CNP.

Alternatively, near real-time seismic activity can be accessed via the NEIC website:

http://earthquake.usgs.gov Page 101 of 236 INFORMATION USE 1 *

• _J

ATTACHMENT 1 EAL Bases This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (QBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-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 plarit. . . . . . . . . . . . . . .

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(2)-0HP-4022-001-007 Earthquake provides the guidance for determining if the OBE earthquake threshold is exceeded and any required response actions. (ref. 2). Because CNP seismic instrumentation does not provide direct and timely indications of having exceeded the OBE ground acceleration, the alternative EAL wording specified in the generic NEI 99-01 HU2 developers note (felt earthquake) is implemented.

CNP Basis Reference(s):

1. FSAR Section 1.3.1 Structures and Equipment

2. 1(2)-0HP-4022-001-007 Earthquake

3. NEI 99-01 HU2 Page 102 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability:

All Definition(s):

PROTECTED AREA - The area encompassed by physical barriers to control access to the .

plant and to the ISFSI.

Basis:

Response actions associated with a tornado on site is provided in 12-0HP-4022-001-010 Severe Weather (ref. 1).

If damage is confirmed visually or by other in-plant indications, the event may be escalated to an ALERT under EAL CA6.1 or SA9.1.

A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an UNUSUAL EVENT regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.

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.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, Sor C.

CNP Basis Reference(s):

1. 12-0HP-4022-001-010 Severe Weather

2. NEI 99-01 HU3 Page 103 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazartjs and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical iso_lation of a SAFETY SYSTEM component needed for the current operating .

mode Mode Applicability:

I I* All

• Definition(s):

• FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or 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 EGGS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

Refer to EALs CA6.1 or SA9.1 for internal flooding affecting one or more SAFETY SYSTEM trains.

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL 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.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, sore.

Page 104 of 236 INFORMATION USE I

I .

ATTACHMENT 1 EAL Bases CNP Basis Reference(s):

1. NEI 99-01 HU3 Page 105 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.3 Unusual Event Movement of personnel within the plant PROTECTED AREA is IMPEDED due to an event external to the PROTECTED AREA involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)

Mode Applicability:

All Definition(s):

/MPEDE(D) - Personnel access to a room or area is hindered to an extent that 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).

PROTECTED AREA - The area encompassed by physical barriers to control access to the plant and to the ISFSI.

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, Sor C.

CNP Basis Reference(s):

1. NEI 99-01 HU3 Page 106 of 236 INFORMATION USE

.J

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.4 Unusual Event A hazardous event that results in onsite conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)

Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

Mode Applicability:

All Definition(s):

None Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL 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.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, Sor C.

CNP Basis Reference(s):

1. NEI 99-01 HU3 Page 107 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):

• 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 The FIRE is located within any Table H-1 area Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table H-1 Fire Areas

• Control Room

• Containment

• Auxiliary Building

• Switchgear Areas

• Diesel Generator Rooms

• ESW System Enclosures

• AFW Pump Rooms

• Refueling Water Storage Tank

• Condensate Storage Tank Mode Applicability:

All Definition(s):

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.

Basis:

Page 108 of 236 INFORMATION USE

---

_______,____J

ATTACHMENT 1 EAL Bases Table H-1 Fire Areas are based on Fire Hazards Analysis Units No. 1 and 2. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1).

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.

For EAL HU4.1 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 th~ 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.

• Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

CNP Basis Reference(s):

1. Fire Hazards Analysis Units No. 1 and 2

2. NEI 99-01 HU4 Page 109 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4- Fire Initiating Condition:* FIRE potentially degrading the level of safety of the plant EAL:

HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE)

AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table H-1 Fire Areas

• Control Room

• Containment

• Auxiliary Building

• Switchgear Areas

• Diesel Generator Rooms

• ESW System Enclosures 0 AFW Pump Rooms

• Refueling Water Storage Tank

• Condensate Storage Tank Mode Applicability:

All Definition(s):

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.

Basis:

Table H-1 Fire Areas are based on Fire Hazards Analysis Units No. 1 and 2. Table H-1 Fire Areas include those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (ref. 1).

Page 110 of 236 INFORMATION USE

• 1

ATTACHMENT 1 EAL Bases 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.

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-rriinute clock starts *at the tinie that the initial alarm was received, a:nd n*ot 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. 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.

If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, arid the emergency must be declared if the FIRE is not extinguished within 15-niinutes 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.

Basis-Related Requirements from Appendix R (Note: CNP is not an Appendix R plant. This bases is cited only to justify the 30 minute timing component related to a single fire alarm)

Appendix 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 explosions."

When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.

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.

In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in this EAL, the 30-minutes to verify a single alarm is well within this worst-case 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.

CNP Basis Reference(s):

Page 111 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases

1. Fire Hazards Analysis Units No. 1 and 2

2. NEI 99-01 HU4 Page 112 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.3 Unusual Event A FIRE within the PROTECTED AREA (plant or ISFSI) not extinguished within 60 min. of the initial report, alarm or indication (Note 1) .

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

. Mode Applicability:

All

• Definition(s):

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.

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.

PROTECTED AREA - The area encompassed by physical barriers to control access to the plant and to the ISFSI.

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.

In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the PROTECTED AREA (plant or ISFSI) 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 an ISFSI located outside the plant PROTECTED AREA.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

CNP Basis Reference(s):

1. NEI 99-01 HU4

. L - ,_ _ _ _ _ _ _ _ _ _ _ _P_ag_e_1_1_3_o_f_23_6_ _ _ _ _ _ _ 1N_F_O_R_M_A_T_IO_N_U_S_E___,

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.4 Unusual Event A FIRE within the PROTECTED AREA (plant or ISFSI) that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability:

All Definition(s):

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.

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.

PROTECTED AREA - The area encompassed by physical barriers to control access to the plant and to the ISFSI.

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.

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.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.

CNP Basis Reference(s):

1. NEI 99-01 HU4

• 1~____________ P_ag_e_1_1_4_o_f_23_6_ _ _ _ _ _ _ 1N_F_O_R_M_A_T_IO_N_U_S_E__.I

' ___J

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gases Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL:**

HA5.1 Alert Release_ of a toxic, corrosive, asphyxiant or flam_mable gas into any Table H-2 rooms or areas AND Entry into the room or _area is prohibited_ or IMPEDED (No_te 5)

. Notei 5: . If_ thei equipment in the listed. room or area Wa$ aJre~dy in9peraqle or out-of7seirvice before the everit occLJrred, then ..

no emergency classification is warranted.

Table H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Applicability Auxiliary Building 573' BART Area 3,4, 5 Auxiliary Building 587' Boric Acid Storage Tank Room, Nuclear 3,4,5 Sampling Room 4KV Room (Mezzanine Area), Boric Acid Batch Tank Area, Chemistry 3, 4, 5 Hot Lab, RHR Hx Room Mode Applicability:

All Definition(s):

IMPEOE(D) - Personnel access to a room or area is hindered to an extent that 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).

Basis:

If the equipment in the listed 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 beyond that already allowed by Technical Specifications at the time of the event.

The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. 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) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).

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 Page 115 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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. **

Evaluation of the IC and EAL do not *require atmospheric sampling; it only re-quires the SEC 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 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.

This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.

Page 116 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases CNP Basis Reference(s):

1. Attachment 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases

2. NEI 99-01 HAS Page 117 of 236 INFORMATION USE *1

.j

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL:*

HA6.1 Alert An event has resulted in plant control being transferred from the Control R.oom to the Local Shutdown. Instrumentation Mode Applicability:

All

• Definition(s):

None Basis:

The Shift Manager (SM) determines if the Control Room is inoperable and requires evacuation.

Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions. 1(2)-0HP-4025-001-001, Emergency Remote Shutdown (ERS), and 1(2)-0HP-4022-CRE-001 Control Room Evacuation provide the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room (Ref. 1 , 2).

Inability to establish plant control from outside the Control Room escalates this event to a SITE AREA EMERGENCY per EAL HS6.1.

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.

Page 118 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases CNP Basis Reference(s):

1. 1(2)-0HP-4025-001-001, Emergency Remote Shutdown (ERS)

2. 1(2)-0HP-4022-CRE-001 Control Room Evacuation

3. NEI 99-01 HA6 Page 119 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:

HS6.1 Site Area Emergency .

An event has resulted in plant control being transferred from the Control Room to the Local Shutdown Instrumentation AND Control of any of the following key safety functions is not reestablished within 15 min.

(Note 1):

• . Reactivity control (modes 1, 2 and 3 only)

• Core cooling

• RCS heat removal Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown, 5 - Cold Shutdown, 6 - Refueling Definition(s):

None Basis:

The Shift Manager determines if the Control Room is inoperable and requires evacuation.

Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions. 1(2)-0HP-4025-001-001, Emergency Remote Shutdown (ERS) and 1(2)-0HP-4022-CRE-001 Control Room Evacuation provides the instructions for tripping the unit, and maintaining RCS inventory and Hot Shutdown conditions from outside the Control Room (ref. 1, 2).

The intent of this EAL is to capture events in which control of the plant cannot be reestablished in a timely manner. The fifteen minute time for transfer starts when the last licensed operator has left the Control Room (not when 1(2)-0HP-4025-001-001 or 1(2)-0HP-4022-CRE-001 is entered). The time interval is based on how quickly control must be reestablished without core uncovery and/or core damage. The determination of whether or not control is established from outside the Control Room is based on SEC judgment. The SEC is expected to make a reasonable, informed judgment that control of the plant from outside the Control Room cannot be established within the fifteen minute interval.

Once the Control Room is evacuated, the objective is to establish control of important plant equipment and maintain knowledge of important plant parameters in a timely manner. Primary Page 120 of 236 INFORMATION USE

- - _J

ATTACHMENT 1 EAL Bases emphasis should be placed on components and instruments that supply protection for and information about safety functions. Typically, these safety functions are reactivity control (ability to shutdown the reactor and maintain it shutdown), RCS inventory (ability to cool the core), and secondary heat removal (ability to maintain a heat sink).

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 can hot 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 Site Emergency Coordinator judgment. The Site Emergency Coordinator 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 FG1 or CG1 CNP Basis Reference(s):

1. 1(2)-0HP-4025-001-001, Emergency Remote Shutdown (ERS)

2. 1(2)-0HP-4022-CRE-001 Control Room Evacuation

3. NEI 99-01 HS6 Page 121 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - SEC Judgment Initiating Condition: Other conditions existing that in the judgment of the SEC warrant declaration of a U E EAL:**

HU7.1 Unusual Event Other cor:iditions exist which in the judgment of the SEC 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.

Mode .Applicability:

All Definition(s):

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 classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Basis:

The SEC is the designated onsite individual having the responsibility and authority for implementing the CNP Emergency Plan (ref. 1). The Shift Manager (SM) initially acts in the capacity of the SEC and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by theSEC, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in

• anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Site Emergency Coordinator to fall under the emergency classification level description for an

• UNUSUAL EVENT.

r Page 122 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases CNP Basis Reference(s):

1. CNP Emergency Plan section 8.5.a.1 Site Emergency Coordinator

2. CNP Emergency Plan section 8.1.k On-Shift Operations Personnel

3. NEI 99-01 HU?

Page 123 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - SEC Judgment Initiating Condition: Other conditions exist that in the judgment of the SEC warrant declaration of an ALERT EAL:

HA7.1 Alert Other conditions exist which, i_n the judgment of the SEC, indicate that events are in progress or have occurred which involve an a.ctual 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 expo_sure_ levels.

Mode Applicability:

All Definition(s):

HOSTILE ACTION - An act toward CNP 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 CNP. 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:

The SEC is the designated onsite individual having the responsibility and authority for implementing the CNP Emergency Plan (ref. 1). The Shift Manager (SM) initially acts in the capacity of the SEC and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the SEC, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Site Emergency Coordinator to fall under the emergency classification level description for an ALERT.

Page 124 of 236 INFORMATION USE I

ATTACHMENT 1 EAL Bases CNP Basis Reference(s):

1. CNP Emergency Plan section 8.5.a.1 Site Emergency Coordinator

2. CNP Emergency Plan section 8.1.k On-Shift Operations Personnel

3. NEI 99-01 HA?

Page 125 of 236 INFORMATION USE J

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - SEC Judgment Initiating Condition: Other conditions existing that in the judgment of the* SEC warrant declaration of a SITE AREA EMERGENCY

• • EAL:

HS7.1 Site.Area Emergency*

Other conditions exist which in the judgment of the SEC 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 boundcfry . . . . . . . . .

Mode Applicability:

All Definition(s):

HOSTILE ACTION - An act toward CNP 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 CNP. 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:

The SEC is the designated onsite individual having the responsibility and authority for implementing the CNP Emergency Plan (ref. 1). The Shift Manager (SM) initially acts in the capacity of the SEC and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the SEC, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Site Emergency Coordinator to fall under the emergency classification level description for a SITE AREA EMERGENCY.

Page 126 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases CNP Basis Reference(s):

1. CNP Emergency Plan section B.5.a.1 Site Emergency Coordinator

2. CNP Emergency Plan section B.1.k On-Shift Operations Personnel

3. NEI 99-01 HS7 Page 127 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - SEC Judgment Initiating Condition: Other conditions exist which in the judgment of the SEC warrant declaration of a GENERAL EMERGENCY EAL:**

HG7.1 General Emergency Other conditions exist which in the judgment of the SEC indicate that events are in .

pr.ogress 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 Mode Applicability:

All Definition(s):

HOSTILE ACTION - An act toward CNP 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 CNP. 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:

The SEC is the designated onsite individual having the responsibility and authority for implementing the CNP Emergency Plan (ref. 1). The Shift Manager (SM) initially acts in the capacity of the SEC and takes actions as outlined in the Emergency Plan implementing procedures (ref. 2). If required by the emergency classification or if deemed appropriate by the SEC, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency.

Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the site boundary.

• This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Site Page 128 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Emergency Coordinator to fall under the emergency classification level description for a GENERAL EMERGENCY.

CNP Basis Reference(s):

1. CNP Emergency Plan section 8.5.a.1 Site Emergency Coordinator

• 2. CNP Emergency Plan section 8.1.k On-Shift Operations Personnel

3. NEI 99-01 HG7

• Page 129 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature> 200°F); EALs in this category are applicable only in one or more hot operating modes.

Numerous system-related equipment failure events that warrant emergency classification have ..

been identified in this category. They may pose actual or potential threats to plant. .

safety.

The events of this category pertain to the following subcategories:

. 1 . Loss of Emergency AC Power Loss of emergency electrical power can compromise plant SAFETY SYSTEM operability .

including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite sources for 4.16KVAC emergency buses: . . . . . . . .

2. Loss of Vital DC Power Loss of emergency electrical power can compromise plant SAFETY SYSTEM operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of vital plant 250 VDC power sources.

3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% - 5%. clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.

5. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and containment integrity.

Page 130 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases

6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor trips. In the plant licensing basis, postulated failures of the RPS to complete a reactor trip comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, A TWS is intet1ded to mean any trip failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and containment integrity.

7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

8. Containment Failure Failure of containment isolation capability {under conditions in which the containment is not currently challenged) warrants emergency classification. Failure of containment pressure control capability also warrants emergency classification.

9. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant SAFETY SYSTEM performance or significant VISIBLE DAMAGE warrant emergency classification under this subcategory.

Page 131 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 minutes or longer EAL:*

SU1.1 Unusual Event Loss of all offsite AC power capability, Table S-1, to emergency 4.16KV buses T11A (T21A) and T11D (T21D) for~ 15 min. (Note 1)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Offsite:

• Unit Auxiliary Xmr TR1AB (TR2AB)

• Unit Auxiliary Xmr TR1 CD (TR2CD)

• Reserve Auxiliary Xmr TR101AB (TR201AB)

• Reserve Auxiliary Xmr TR101 CD (TR201 CD)

• 69/4.16 kV Alternate XmrTR12EP-1 Onsite:

• EOG 1AB (2AB)

• EOG 1CD (2CD)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

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.

A list of offsite AC power sources credited for this EAL are specified in Table S-1.

The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. 4.16KV buses T11A (T21A) and T11 D (T21 D) are the emergency (essential) buses (ref. 1). While generating, auxiliary power is normally supplied from the generator terminals through the unit auxiliary transformers (TR1AB and TR1 CD for Unit 1 and TR2AB and TR2CD for Unit 2). When the plant trips or the plant is shutdown the station auxiliaries are transferred to the preferred offsite power source (that is, to reserve auxiliary transformers TR 101 AB and TR1 otco for Unit 1 and TR201AB and TR201 CD for Unit 2) to assure contin.ued power to equipment when the main generator is off-line (ref. 1, 2, 3).

Page 132 of 236 INFORMATION USE I

_._ _.__J

ATTACHMENT 1 EAL Bases In addition, an alternate offsite power source, a 69/4.16kV transformer (TR 12EP-1 ), located at the plant site, has the necessary capacity to operate one train of the engineered safeguard equipment in one unit while supplying one train of the safe shutdown power in the other.

T11A {T21A) and T11 D (T21 D) also each have an emergency diesel generator which supply onsite electrical power to the bus automatically in the event that the preferred offsite sources

• become unavailable (ref. 1, 2, 3). * * **

The Supplemental Diesel Generators (SDGs) *are not credited a*s an AC power source for this .

EAL.

a This IC addresses 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, "capabilityn.means that an offsite AC power source(s) is available to the.emergency buses, whether or not the buses are.powered from it. .

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 SA 1.

CNP Basis Reference(s):

1. UFSAR Figure 8.1-1A(B) Main Auxiliary One-Line Diagram

2. UFSAR Section 8.0 Electrical Systems

3. 1(2)-0HP-4022-001-005 Loss of Offsite Power with Reactor Shutdown

4. NEI 99-01 SU1 Page 133 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL:

SA1.1 Alert AC power capability, Table S-1, to emergency 4.16KV buses T11A (T21A) and T11 D (T21D) reduced to a single power source for;:: 15 min. (Note 1)

AND Any additional s_ingle power s_ource fai_lure will result in l9ss of all_ AC power to SAFETY SYSTEMS Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table S-1 AC Power Sources Offsite:

• Unit Auxiliary Xmr TR1AB (TR2AB)

• Unit Auxiliary Xmr TR1 CD (TR2CD)

• Reserve Auxiliary Xmr TR101AB (TR201AB)

• Reserve Auxiliary Xmr TR101CD (TR201CD)

• 69/4.16 kV Alternate Xmr TR12EP-1 Onsite:

• EDG 1AB (2AB)

• EDG 1CD (2CD)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

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 classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Page 134 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Basis:

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.

A list of onsite and offsite AC power sources credited for this EAL are specified in Table S-1.

The 4.16KV AC System provides the power requirements for operation and safe shutdown of

• the plant. 4.16KV buses T11 A(T21 A) and T11 D (T21 D) are the emergency (essential) buses

• (ref. 1). While generating, auxiliary power is normally supplied from the generator terminals through the unit auxiliary transformers (TR1AB and TR1CD for Unit 1 and TR2AB and TR2CD for Unit 2). Whe_n the plant trips or the piant is shutdown the station auxiliaries are transferred to the preferred offsite power source (that is, to reserve auxiliary transformers TR101AB and TR101 CD for Unit 1 and TR201AB and TR201 CD for Unit 2) to assure continued power to e_quipmen_t when the main generator is off-line (ref. 1, 2, 3) .

.. In addition, an alternate offsite power source, a 69/4.16kV transformer (TR12EP-1), located at the plant site, has the necessary capacity to operate one train of the engineered safeguard equipment in one unit while supplying one train of the safe shutdown power in the other.

T11A (T21A) and T11 D (T21 D) also each have an emergency diesel generator which supply onsite electrical power to the bus automatically in the event that the preferred offsite sources become unavailable (ref. 1, 2, 3).

The Supplemental Diesel Generators (SDGs) are not credited as an AC power source for this EAL.

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 SU1.

An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Examples of this 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 emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being 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 SS1.

CNP Basis Reference(s):

1. UFSAR Figure 8.1-1 A(B) Main Auxiliary One-Line Diagram

2. UFSAR Section 8.0 Electrical Systems

3. 1(2)-0HP-4022-001-005 Loss of Offsite Power with Reactor Shutdown

4. NEI 99-01 SA1 Page 135 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to emergency buses for 15 minutes or longer EAL:

SS1.1 *site Area Emergency Loss of all offsite and all onsite AC power to emergency 4.16KV buses T11A (T21A) and Tt 1D (T21 D) for 2: 1.5 min. (Note 1)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

• Mode Applicability:

• 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. 4.16KV buses T11A (T21 A) and T11 D (T21 D) are the emergency (essential) buses (ref. 1). While generating, auxiliary power is normally supplied from the generator terminals through the unit auxiliary transformers (TR1AB and TR1 CD for Unit 1 and TR2AB and TR2CD for Unit 2). When the plant trips or the plant is shutdown the station auxiliaries are transferred.

to the preferred offsite power source (that is, to reserve auxiliary transformers TR101AB and TR101 CD for Unit 1 and TR201AB and TR201 CD for Unit 2) to assure continued power to equipment when the main generator is off-line (ref. 1, 2, 3).

In addition, an alternate offsite power source, a 69/4.16kV transformer (TR12EP-1 ), located at the plant site, has the necessary capacity to operate one train of the engineered safeguard equipment in one unit while supplying one train of the safe shutdown power in the other.

T11A (T21A) and T11 D (T21 D) also each have an emergency diesel generator which supply onsite electrical power to the bus automatically in the event that the preferred offsite sources become unavailable (ref. 1, 2, 3).

The Supplemental Diesel Generators (SDGs) or any other alternative AC power source

_capable of powering an emergency bus can also be credited as an AC power source for this .

EAL.

The 15-minute interval begins when both offsite and onsite AC power capability are lost (ref.

4).

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 contml, spent fuel ~eat removal and the ultimate heat sink ..

In addition, fission product barrier monitoring capabilities may be degraded under these Page 136 of 236 INFORMATION USE I .

ATTACHMENT 1 EAL Bases conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of 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 ICs RG1, FG1 or SG1.

CNP Basis Reference(s):

1. UFSAR Figure 8.1-1A(B) Main Auxiliary One-Line Diagram *

2. UFSAR Section 8.0 Electrical Systems

3. 1(2)-0HP-4022-001-005 Loss of Offsite Power with Reactor Shutdown

4. 1(2)-0HP-4023-ECA-O.O Loss of All AC Power 5 .. NEI 99-01 SS1

.. ~'~~~~~~~~~~~-P_a_g_e_1_3_7_o_f_23_6~~~~~~~1N~FO~R_M_A_T_IO_N~U_S_E---J

ATTACHMENT 1 EAL Bases Category: S -System Malfunction Subcategory: 1 - Loss of Emergen~y AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses EAL:

SG1 .1 General Emergency Loss of all offsite and all on site AC power to emergency 4.16KV buses T11 A (T21 A) and T11 D (T21 D) * .

AND EITHER:

• . Restoration of at_ least one emergency bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1)

• CSFST Core Cooling REQ Path (F.0-2) conditions met Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

. *, . -~., .,: , '::-. .' '

. 1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

This EAL is indicated by the extended loss of all offsite and onsite AC power capability to 4.16KV emergency buses T11A (T21A) and T11D (T21 D) either for greater then the CNP Station Blackout (SBO) coping analysis time (4 hrs.) (ref. 1, 4) or that has resulted in indications of an actual loss of adequate core cooling.

Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met. (ref. 5).

The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. 4.16KV buses T11A (T21A) and T11 D (T21 D) are the emergency (essential) buses (ref. 1). ,While generating, auxiliary power is normally supplied from the generator terminals through the unit auxiliary transformers (TR1AB.and TR1 CD for Unit 1 and TR2AB and TR2CD for Unit 2). When the plant trips or the plant is shutdown the station auxiliaries are transferred to the preferred offsite power source (that is, to reserve auxiliary transformers TR101AB and TR101 CD for Unit 1 and TR201AB and TR201 CD for Unit 2) to assure continued power to equipment when the main generator is off-line (ref. 1, 2, 3).

In addition, an alternate offsite power source, a 69/4.16kV transformer (TR12EP-1 ), located at the plant site, has the necessary capacity to operate one train of the engineered safeguard

• equipment in one unit while supplying one train* of the safe shutdown power in the other.

.j Page 138 of 236 . INFORMATION USE I~

ATTACHMENT 1 EAL Bases T11A (T21A) and T11 D (T21 D) also each have an emergency diesel generator which supply onsite electrical power to the bus automatically in the event that the preferred offsite sources become unavailable (ref. 1, 2, 3).

The Supplemental Diesel Generators (SDGs) or any other alternative AC power source capable of powering an emergency bus can also be credited as an AC power source for this EAL.**

Four hours is the station blackout coping tinie (ref. 4) ..

Indication of continuing core cooling degradation must be based on fission product barrier monitoring with. particular emphasis o.n SEC judgment as it relates to IMMINENT loss of fission product barriers and degraded ability to monitor fission product barriers. Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH condmons being met. Specifically, Core CooHng RED PATH conditions exist if either.the ~ve .

highest core exit TCs are reading greater than or equal to 1200°F or core exit TCs are reading greater than orequal to 757°F with RCS subccioling less than or'equal to 40°F, arid RVLIS.

indication is less than or equal to that specified based on the number of RCPs running (ref. 5).

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.

The EAL should require declaration of a GENERAL EMERGENCY prior to meeting the thresholds for IC FG1. 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.

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 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.

CNP Basis Reference(s):

1. UFSAR Figure 8.1-1A(B) Main Auxiliary One-Line Diagram

2. UFSAR Section 8.0 Electrical Systems

3. 1(2)-0HP-4023-ECA-O.O Loss of All AC Power

4. UFSAR Section 8.7 Station Blackout .

5. 1(2)-0HP-4023-F-0.2 Core Cooling Page 139 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases

6. NEI 99-01 SG1 Page 140 of 236 INFORMATION USE

- -- ____ ,____c_J

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 2 - Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer

• EAL:

SS2.1 Site Area Emergency Loss of all 250 VDC power based on bus voltage indications < 215 VDC on all vital DC buses 1CD (2CD) (Train A) and 1AB (2AB) (Train B) f9r 2: 15 min. (Note 1)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

. Mode Applicability: .

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 -.Hot Shutdown Definition(s):

None Basis:

The vital DC buses are the following 250 VDC Class 1E buses (ref. 1, 2, 3):

Train A: Train B:

1CD (2CD) 1AB (2AB)

There are two, 116 cell, lead-acid storage batteries (1AB (2AB) and 1CD (2CD)) that supplement the output of the battery chargers. They supply DC power to the distribution buses when AC power to the chargers is lost or when transient loads exceed the capacity of the battery chargers (ref. 3).

CNP Technical Specification LCO 3.8.4 requires that both Train A and Train B 250 VDC electrical power subsystem shall be OPERABLE to support both trains of the DC Electrical Power Distribution System required by LCO 3.8.9, "Distribution Systems - Operating." (ref. 1).

Per SD-DCC-NEEP-104, a 210 VDC lower limit has been identified from the battery service test acceptance criteria. Based on interpolation, the low voltage limit that would provide a 15 minute margin has been determined to be 213 VDC (ref. 4).

An EAL value of 215 VDC has been selected to account for available instrument accuracy.

Meter scaling on installed control room instrumentation (10 VDC divisions on a dial indicator) limits the closest value that can be accurately read on the control board to 5 VDC.

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 of 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 ICs RG1, FG1 or SG2.

Page 141 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases CNP Basis Reference(s):

1. Technical Specifications Section 3.8.4 DC Sources - Operating

2. UFSAR Figure 8.3-2

3. UFSAR Section 8.3.4 250 Volt DC System (Safety Related)

4. SD-DCC-NEEP-104 250 voe System

5. NEI 99-01 SS8 Page 142 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S -System Malfunction Subcategory: 2 - Loss of Vital DC Power Initiating Condition: Loss of all AC and vital DC power sources for 15 minutes or longer*

EAL:

SG2.1 . General Emergency.

Loss of all offsite and all on site AC power to emergency 4.16KV buses T11 A (T21 A) and T11 D (T21 D) for~ 15 min:

AND Loss of all 250 VDC power based on bus voltage indications < 215 VDC on all vital DC buse.s 1CD (2CD). (Train A) and 1AB (2AB) (Train B) for~. 15 min ..

(Note 1).

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

This EAL is indicated by the loss of all offsite and onsite emergency AC power capability to 4.16KV emergency buses T11A (T21A) and T11 D (T21 D) for greater than 15 minutes in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.

The 4.16KV AC System provides the power requirements for operation and safe shutdown of the plant. 4.16KV buses T11A (T21A) and T11 D (T21 D) are the emergency (essential) buses (ref. 1). While generating, auxiliary power is normally supplied from the generator terminals through the unit auxiliary transformers (TR1AB and TR1 CD for Unit 1 and TR2AB and TR2CD for Unit 2). When the plant trips or the plant is shutdown the station auxiliaries are transferred to the preferred offsite power source (that is, to reserve auxiliary transformers TR101AB and TR101 CD for Unit 1 and TR201AB and TR201 CD for Unit 2) to assure continued power to equipment when the main generator is off-line (ref. 1, 2, 3).

In addition, an alternate offsite power source, a 69/4.16kV transformer (TR12EP-1), located at the plant site, has the necessary capacity to operate one train of the engineered safeguard equipment in one unit while supplying one train of the safe shutdown power in the other.

Page 143 of 236 INFORMATION USE L.

ATTACHMENT 1 EAL Bases T11A (T21A) and T11 D (T21 D) also each have an emergency diesel generator which supply onsite electrical power to the bus automatically in the event that the preferred offsite sources become unavailable (ref. 1, 2, 3).

The Supplemental Diesel Generators (SDGs) or any other alternative AC power source capable of powering an emergency bus can also be credited as an AC power source for this EAL. *.

The vital DC buses are the following 250 VDC Class 1E buses (ref. 4, 5, 6):

Train A: Train B:

1 en (2CD) . 1AB (2AB)

There are two, 116 cell, lead-acid storage batteries (1 AB (2AB) and 1CD (2CD)) that supplement the output of the battery chargers. They supply DC power to the distribution buses

. when AC power to the chargers is lost or when transient loads exceed the capacity of the batte.ry chargers (ref. >).

CNP Technical Specification LCO 3.8.4 requires that both Train A and Train B 250 VDC electrical power subsystem shall be OPERABLE to support both trains of the DC Electrical Power Distribution System required by LCO 3.8.9, "Distribution Systems - Operating." (ref. 4).

Per SD-DCC-NEEP-104, a 210 VDC lower limit has been identified from the battery service test acceptance criteria. Based on interpolation, the low voltage limit that would provide a 15 minute margin has been determined to be 213 VDC (ref. 7).

An EAL value of 215 VDC has been selected to account for available instrument accuracy.

Meter scaling on installed control room instrumentation (10 VDC divisions on a dial indicator) limits the closest value that can be accurately read on the control board to 5 VDC.

This IC addresses a concurrent and prolonged loss of both emergency AC and Vital DC power. A loss of all emergency 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 emergency AC and vital DC power will lead to multiple challenges to fission product barriers.

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.

CNP Basis Reference(s):

1. UFSAR Figure 8.1-1A(B) Main Auxiliary One-Line Diagram

2. UFSAR Section 8.0 Electrical Systems

3. 1(2)-0HP-4023-ECA-O.O Loss of All AC Power Page 144 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases

4. Technical Specifications Section 3.8.4 DC Sources - Operating

5. UFSAR Figure 8.3-2

6. UFSAR Section 8.3.4 250 Volt DC System (Safety Related)

7. SD-DCC-NEEP-104 250 voe System

8. NEI 99-01 SG8 Page 145 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:

SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for;?: 15 min. (Note 1)

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

7 aule S-2 *Safety System Parameters

• Reactor power

• RCS level

• RCS pressure

• Core Exit TC temperature

• Level in at least one SG

• Auxiliary feed flow in at least one SG Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

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:

SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.

The Plant Process Computer, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1).

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 parametet(s). For example, the reactor Page 146 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 NRG event report is required. The event would be reported if it significantly impaired the

• capability to perform emergency assessments. In particular, emergency assessme*nts **

.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 IC SA3.

CNP Basis Reference(s):

1. UFSAR Section 7.5 Engineered Safety Features Instrumentation

2. NEI 99-01 SU2 Page 147 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:

SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S"'.2 parameters from within the.Control Room for;:: 15 min. (Note 1)

AND Any sig!lificant transient is in progress, Table S-:-3 Note 1:

• The SEC should declare the event promptly upon determining that time limit has been exceeded, or will*

likely be exceeded.

'!°""hie ~-2 ~ .. fety System Parameters

• Reactor power

• RCS level

• RCS pressure

• Core Exit TC temperature

• Level in at least one SG

• Auxiliary feed flow in at least one SG

• Reactor trip

• Runback;:: 25% thermal power

• Electrical load rejection > 25%

of full electrical load

• ECCS actuation

• Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

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.

Page 148 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Basis:

SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems.

The Plant Computer, which displays SPDS required information, serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators (ref. 1).

Significant transients are listed in Table S-3 and include response to automatic or manually initiated functions such as reactor trips, runbacks involving greater than or equal to 25% .

thermal power change, electrical load rejections of greater than 25% full electrical load or ECCS (SI) injection actuations.

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 C_ontrol Room. Owing this condition, the margin to a potential fissio_n produ~t barrier challenge is reduced. It thus represents a potential substantial 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.

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 FS1 or IC RS1 CNP Basis Reference(s):

1. UFSAR Section 7.5 Engineered Safety Features Instrumentation

2. NEi 99-01 SA2 Page 149 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 4- RCS Activity Initiating Condition:

• Reactor coolant activity greater than Technical Specification allowable limits

• EAL:

SU4.1 unusual Event Sample analysis indicates RCS activity > Tec.hnical Specification Section 3.4.16 limits Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. 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 FA1 or the Recognition Category R ICs.

CNP Basis Reference(s):

1. CNP Technical Specifications section 3.4.16 RCS Specific Activity

2. NEI 99-01 SU3 Page 150 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL:

SU5.1 Unusual Event RCS unidentified or pressure boundary leakage> 10 gpm for;?; 15 min.

OR RCS identified leakage > 25 gpm for;?; 15 min.

OR Leakag*e from ttie RCS to a location outsid*e contairimerit > 25 gpm for;?; 15 m*in.

(Note 1) --

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

Manual or computer-based methods of performing an RCS inventory balance are normally used to determine RCS leakage (ref. 1).

Identified leakage includes

• Leakage such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank, or

• Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage, or

• RCS leakage through a steam generator to the secondary system (ref. 2).

- Unidentified leakage is all leakage (except RCP seal water injection or leakoff) that is not identified leakage (ref. 2).

Pressure Boundary leakage is leakage (except SG leakage) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall (ref. 2)

RCS leakage outside of the containment that is not considered identified or unidentified

. leakage per Technical Specifications includes leakage via interfacing systems such as RCS to the Component Cooling Water,-or systems that directly see RCS pressure outside containment Page 151 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases such as Chemical & Volume Control System, Nuclear Sampling system and Residual Heat Removal system (when in the shutdown cooling mode) (ref. 3, 4)

Escalation of this EAL to the ALERT level is via Category F, Fission Product Barrier Degradation, EAL FA1 .1.

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.

a The first and second EAL conditions are focused on 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). The third condition addresses an RCS mass .loss.caused by ar:i UN.ISOLABLE leak through an interfacing system. These co.nditions .

thus apply to leakage into the containment, a secondary-side system (e.g., steam generator

. tube leakage) or*a location outside of cont~iinment.

The leak rate values for each condition 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). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. 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.

CNP Basis Reference(s):

1. 1(2)-0HP-4030-102-016 Reactor Coolant System Leak Test

2. CNP Technical Specifications Definitions section 1.1

3. UFSAR Section 4.2.7 Leakage

4. 1(2)-0HP-4022-002-020 Excessive Reactor Coolant Leakage

5. NEI 99-01 SU4 Page 152 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:

SU6.1 Unusual Event An automatic trip did not shut down the reactor as indicated by reactor power~ 5% after any RPS setpoint is exceeded AND A subsequent automatic trip or manual trip action taken at the reactor control console (reactor trip switches) is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8) . . . . . . . . . . . . . . .

Note 8: A manual trip 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.

Mode Applicability:

1 - Power Operation Definition(s):

None Basis:

The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) trip function. A reactor trip is automatically initiated by the RPS when certain continuously monitored parameters exceed predetermined setpoints (ref. 1, 2).

Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from an automatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful trip has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2, 3, 4).

For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; manual reactor trip switches. Reactor shutdown achieved by use of other trip actions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as tripping the main turbine, locally opening reactor trip breakers, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).

Following any automatic RPS trip signal, E~O (ref. 2) and /FR-s.1 * (ref. 4) prescribe insertion of redundant manual trip signals to back up the automatic RPS trip function and ensure reactor Page 153 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases shutdown is achieved. Even if the first subsequent manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the automatic trip, the lowest level of classification that must be declared is an UNUSUAL EVENT (ref. 4).

A reactor trip resulting from actuation of the ATWS Mitigation System Actuation Circuitry (AMSAC) logic that results in full insertion of control rods and diminishing neutron flux is

• considered a successful reactor trip. AMSAC automatically initiates auxiliary feedwa:ter and a

• turbine trip under conditions indicative.of an Anticipated Transient Without Scram (ATWS) ..

event (ref. 5).

• In the event that the operator identifies a reactor trip is IMMINENT and initiates a successful manual reactor trip before the automatic RPS trip setpoint is reached, no declaration is required. The successful manual trip of the reactor before it reaches its automatic trip setpoint or reactor trip signals caused by instrumentation channel failures do not lead to a potential

• fission product barrier loss. However, if subsequent manual reactor trip actions*fail to reduce

• reactor power below 5%, the event escalates to the ALERT under EAL SA6.1.

If by procedure, operator actions include the initiation of an immediate manual trip following receipt of an automatic trip signal and there are no clear indications that the automatic trip failed (such as a time delay following indications that a trip setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic trip or manual actions.

If a subsequent review of the trip actuation indications reveals that the automatic trip did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50.72 should be considered for the transient event.

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.

Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles 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 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. Actions taken at back-panels or other lo.cations within the Control Room,*

or ariy location outside the Control Room, are not considered to be "at the reactor control consoles".

Page 154 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 FA 1. Absent the plant conditions needed to meet either IC SA6 or FA 1, an UNUSUAL EVENT dedaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Shou_ld a _reactor trip signal be generated as a result of pla_nt 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.

CNP Basis Reference(s):

1. CNP Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation

2. 1(2)-0HP04023-E-O Reactor Trip or Safety Injection

3. 1(2)-0HP04023-F-0.1 Critical Safety Function Status Trees - Subcriticality

4. 1(2)-0HP-4023-FR-S-1 Response to Nuclear Power Generation/ATWS

5. UFSAR Section 3.3.3 Anticipated Transients Without Scram 6 NEI 99-01 SUS Page 155 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual trip fails to shut down the reactor EAL:

SU6.2* Unusual Event A manual trip did not shut down the reactor as indicated by reactor power;?! 5% after any manual trip adion was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (reactor trip switches) is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8) . - - . - . - . . - - . - - .

Note 8: A manual trip 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.

Mode Applicability:

1 - Power Operation Definition(s):

None Basis:

This EAL addresses a failure of a manually initiated trip in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual trip is successful in shutting down the reactor (reactor power< 5%). (ref. 1).

Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative startup rate. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-trip response from an automatic reactor trip signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful trip has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the immediate shutdown decay heat level of 5% (ref. 2, 3, 4).

For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; manual reactor trip switches. Reactor shutdown achieved by use of other trip actions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as tripping the main turbine, locally opening reactor trip breakers, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).

Following the failure of any manual trip signal, E-0 (ref. 2) and FR-S.1 (ref. 4) prescribe insertion ofredundant manual trip signals to back up the RPS trip function and ensure reactor shutdown is achieved. Even if a subsequent automatic trip signal or the first subsequent Page 156 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases manual trip signal inserts all control rods to the full-in position immediately after the initial failure of the manual trip, the lowest level of classification that must be declared is an UNUSUALEVENT(ref~.

A reactor trip resulting from actuation of the A TWS Mitigation System Actuation Circuitry (AMSAC) logic that results in full insertion of control rods and diminishing neutron flux is considered a successful reactor trip. AMSAC automatically initiates auxiliary feedwater and a turbine trip under conditions indicative of an Anticipated Transient Without Scram (ATWS) .

event (ref. 5).

• If both subsequent automatic and subsequent manual reactor trip actions in the Control Room fail to reduce reactor power below the power associated with the SAFETY SYSTEM design(<

5%) following a failure of an initial manual trip, the event escalates to an ALERT under EAL SA6.1.

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor a

trip that results in reactor'shutdown, arid eithera slibsec:juerit 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.

Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles 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 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. 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 FA 1. Absent the plant conditions needed to meet either IC SA6 or FA 1; an

• UNUSUAL EVENT declaration is appropriate for this event.

• A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

• 1 Page 157 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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., assessm*ent of test results), then this IC and the EALs are riot applicable and no classification is warranted.

CNP Basis Reference(s):

1. CNP Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation

2. 1(2)-0HP94023-E~O Reactor Trip or _Safety l~jection 3 .. 1(2)-0HP04023-:F-0,.1 Critical Safety Function Status Trees -: Sut:>criticality

4. 1(2)-0HP-4023-FR-S-1 Response to Nuclear Power Generation/ATWS

5. UFSAR Section 3.3.3 Anticipated Transients Without Scram

6. NEI 99-01 SUS

• Page 158 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 2 - RPS Failure

• Initiating Condition: Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control console are not successful in shutting down the reactor EAL:

SA6.1 Alert An automatic or manual trip fails to shut down the reactor as indicated by reactor power

5%

AND Manual trip actions taken a*t the reactor control console (reactor trip swHches) are not successful in shutting down the reactor as indicated by reactor power;;:: 5% (Note 8)

Note 8: A manual trip 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.

Mode Applicability:

1 - Power Operation Definition(s):

None Basis:

This EAL addresses any automatic or manual reactor trip signal that fails to shut down the reactor (reactor power< 5%) followed by a subsequent manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed (ref. 1, 2).

For the purposes of emergency classification, successful manual trip actions are those which can be quickly performed from the reactor control console; manual reactor trip switches. Reactor shutdown achieved by use of other trip actions specified in FR-S.1 Response to Nuclear Power Generation/ATWS (such as tripping the main turbine, locally opening reactor trip breakers, emergency boration or manually driving control rods) do not constitute a successful manual trip (ref. 4).

A reactor trip resulting from actuation of the ATWS Mitigation System Actuation Circuitry (AMSAC) logic that results in full insertion of control rods and diminishing neutron flux is considered a successful reactor trip. AM SAC automatically initiates auxiliary feedwater and a turbine trip under conditions indicative of an Anticipated Transient Without Scram (ATWS) event (ref. 5).

5% rated power is a minimum reading on the power range scale that indicates continued

power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below 5%, plant response will be similar to that observed during a Page 159 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than 5 % power (ref. 3, 4).

Escalation of this event to a SITE AREA EMERGENCY would be under EAL SS6.1 or SEC judgment.

. 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 ifthe reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS .

. A manual action_ at the reacto.r control .console i.s 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 niamially driving in *control fods or implementatiori" of boron*

injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at backpanels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the reactor control console".

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 shut down 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 possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, 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.

CNP Basis Reference(s):

1. CNP Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation

2. 1(2)-0HP04023-E-O Reactor Trip or Safety Injection

3. 1(2)-0HP04023-F-0.1 Critical Safety Function Status Trees - Subcriticality

4. 1(2)-0HP-4023-FR-S-1 Response to Nuclear Power Generation/ATWS

5. UFSAR Section 3.3.3 Anticipated Transients Without Scram

6. NEI 99-01 SAS Page 160 of 236 INFORMATION USE

• 1

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 2 - RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to core cooling or RCS heat removal

• EAL:

SSS.1 Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power

so/o ..

AND All actions to shut down the reactor are not successful as indicated by reactor power .

so/o AND EITHER

• CSFST Core Cooling.RED Path (F-0.2) conditions met

• CSFST Heat Sink RED Path (F-0.3) conditions met

• • Mode Applicability:

1 - Power Operation Definition(s):

None Basis:

This EAL addresses the following:

• Any automatic reactor trip signal followed by a manual trip that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the SAFETY SYSTEMS were designed (EAL SA6.1 ), and

• Indications that either core cooling is extremely challenged or heat removal is extremely

. challenged. *'

The combit1ation of failure* of both front lihe and backup protection systems to function in.

response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers.

Reactor shutdown achieved by use of FR-S.1 Response to Nuclear Power Generation/ATWS

• s.uch as tripping. 'the main turbine, locally.opening reactor trip breakers, emergency boration or manually driving control rods are also credited as a successful means of shutting down the reactor provided reactor power can be reduced below 5% before indications of an extreme challenge to either core cooling or heat removal exist (ref. 1, 2).

5% rated power is a minimum reading on the power range scale that indicates continued power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core .damage. Below 5%, plant response will be similar to that observed during a j*

• Page 161 of 236 INFORMATION USE . J

ATTACHMENT 1

• EAL Bases normal shutdown. Nuclear instrumentation can be used to determine if reactor power is greater than 5 % power (ref. 1, 2).

Indication of continuing core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met. Specifically, Core Cooling RED PATH conditions exist if either the five highest core exit TCs are reading greater than or equal to 1200°F or core exit TCs are reading greater than or equal to 757°F with RCS subcooling less than or equal 40°F and **

RVLIS level less than or equal to that specified based on the number of RCPs running (ref. 3).

Indication of inability to adequately remove heat from the RCS is manifested by CSFST Heat Sink RED PATH conditions being met (ref. 2). Specifically, Heat Sink RED PATH conditions*

exist if narrow range level in at least one steam generator is not greater than 13% (28%

Adverse Containment Conditions) and total feedwater flow to the steam generators is less than or equal to 240,000 lbm/hr. (ref. 4).

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 shut down 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 RG1 or FG1.

CNP Basis Reference(s):

1. 1(2)-0HP04023-F-0.1 Critical Safety Function Status Trees - Subcriticality

2. 1(2)-0HP-4023-FR-S-1 Response to Nuclear Power Generation/ATWS

3. 1(2)-0HP04023-F-0.2 Critical Safety Function Status Trees - Core Cooling

4. 1(2)-0HP04023-F-0.3 Critical Safety Function Status Trees - Heat Sink

5. NEI 99-01 SS5 Page 162 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

SU7.1 Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 ORO communication methods OR Loss of ali Table S~4 NRC communication methods Table S-4 Communication Methods System Onsite ORO NRC Plant Page X Plant Radios X X Plant Telephone X X X ENS Line X X Commercial Telephone X X Microwave Transmission X X Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

Onsite/offsite communications include one or more of the systems listed in Table C-5 (ref. 1).

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 OROs 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-Page 163 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.).

The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations.

The second EAL condition addresses a total loss of the communications methods used to .

notify all OROs of an emergency declaration. The OROs referred to here are the State and Berrien County EOCs .

The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergericy declaration.. * * * *

• CNP Basis Reference(s):

1. CNP Plant Emergency Plan Section F Emer~ency Communication.s

2. NEI 9£1-01 SU6.

Page 164 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S - System Malfunction Subcategory: 8 - Containment Failure Initiating Condition: Failure to isolate containment or loss of containment pressure control.

EAL:

SU8.1 Unusual Event Any penetration is not isolated within 15 min. of a VALID containment isolation signal OR Containment pressure > 2.8 psig with < one full train of containment depressurization equipment operating per design for;:: 15 min. (Note 9)

• (Note 1)*

Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 9: One Containment Spray System train and one Containment Air Recirculation Fan comprise one full train of depressurization equipment.

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

Basis:

The containment isolation system provides the means of isolating the various pipes passing through the containment walls as required to prevent the release of radioactivity to the outside environment in the event of a design basis accident (ref. 1).

Containment pressure control is achieved through the Containment Spray System and the Containment Air Recirculation/Hydrogen Skimmer System. Failure of either of these systems may allow steam to build up within containment, and, unabated, this steam buildup may cause the internal containment pressure buildup to exceed the design pressure of 12 psig. Studies have shown that the containment can withstand pressures well above this value.

Both the recirculation fans and the containment spray pumps are actuated automatically (time delayed) following receipt of a HI or HI HI (Phase B) containment pressure *signal, respectively.

Since the HI HI containment pressure setpoint is less than or equal to 2.8 PSI, then greater than 2.8 PSI would be the containment pressure greater than the setpoint at which the equipment was_suppos~d to t,ave actuate_d per design. If these systems should fail to start automatically per design, a successful nianual start within 15 minutes would preclude .

exceeding this Containment Potential Loss threshold. (ref. 2, 3, 4)

Page 165 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases This EAL 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 the first condition, the contairimerit isolation signal must be generated as the result of an

• off-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 th.e plant AOPs and EOPs. This condition includes the failure of Containment Ventilation Isolation to actuate on a VALID signal. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible ..

. The second condition 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. 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 (e.g., containment sprays or containment recirculation fans) are either lost or performing in a degraded manner.

This event would escalate to a SffE AREA EMERGENCY in accordance with IC FS1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.

CNP Basis Reference(s):

1. UFSAR Section 5.4 Containment Isolation System

2. UFSAR Section 5.5.3 System Description

3. UFSAR Section 6.3 Containment Spray Systems

4. EC-0000052930 Unit 1 Return to Normal Operating Pressure and Temperature (NOP/NOT)

5. NEI 99-01 SU7 Page 166 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: S- System Malfunction Subcategory: 9 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode EAL:

SA9.1 Alert The occurrence of any Table S-5 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the_ current operating mode AND .EITHER:

• Event damage has caused indications of degraded performance to the second train of the SAFETY SYSTEM needed for the current operating mode

• Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode (Notes 11, 12)

Note 11: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 12: 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.

Table S-5 Hazardous Events

• Seismic event (earthquake)

• Internal or external FLOODING event

• High winds or tornado strike

• FIRE

• EXPLOSION

• Other events with similar hazard characteristics as determined by the SEC Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

Page 167 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 require a post-event inspection to determine if the attributes of an explosion are present.

• FIRE - C6mbustioh 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.

FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutd9wn condition, including the ECCS .. These are typica_lly systems .

classified as safety-related (as defined in 10CFR50.2):

Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

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.

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; commercial nuclear power plants are designed to be able to support single system issues I

I without compromising public health and safety from radiological events.

I 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 Page 168 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases 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 FS1 or RS1.

CNP Basis Reference(s):

1. EP FAQ 2016-002

2. NEI 99-01 SA9 Page 169 of 236 INFORMATION USE

• . _J

ATTACHMENT 1 EAL Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200°F); EALs in this category are applicable only in one or more hot operating modes.

EALs in this category represent threats to the defense in depth design concept that precludes the release of highly 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:

A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.

B. Reactor Coolant System (RCS): The RGS B?rrier incl_udes 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. . .

C. Containment (CNMT): 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.

The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1 (Attachment 2). "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level:

Alert:

Any Joss or any potential Joss of either Fuel Clad or RCS Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

Loss of any two barriers and loss or potential Joss of third barrier The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:

• The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Containment Barrier.

-* UNUSUAL EVENT ICs associated with RCS and Fuel Clad Barriers are addressed under System Malfunction ICs.

• For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to I Page 170 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a SITE AREA EMERGENCY classification while a dose assessment may indicate that an EAL for GENERAL EMERGENCY IC RG1 has been exceeded.

• The fission product barrier thresholds specified within a scheme reflect plant-specific CNP design and operating characteristics.

• As used-in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location- inside the primary containment, an interfacing system, or outside of the primary containment. The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage .

.* At the SITE AREA EMERGENCY level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a GENERAL EMERGEN-CY declaration. For example, if the Fuei Clad and RCS fission product" barriers -

were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the SEC would have more assurance that there was no immediate need to escalate to a GENERAL EMERGENCY.

Page 171 of 236 INFORMATION USE

_j

ATTACHMENT 1 EAL Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL:

FA1.1 Alert Any loss or any potential loss of EITHER Fuel Clad OR RCS (Table F-1)

Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the ALERT classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a SITE AREA EMERGENCY under EAL FS1 .1.

CNP Basis Reference(s):

1. NEI 99-01 FA1 Page 172 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL:

FS1 .1 Site Area Emergency Loss or potential loss of any two barriers (Table F-1)

. Mode Applicability:

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the SITE AREA EMERGENCY classification level, each barrier is weighted equally. A SITE AREA EMERGENCY is therefore appropriate for any combination of the following conditions:

• One barrier loss and a second barrier loss (i.e., loss - loss)

• One barrier loss and a second barrier potential loss (i.e., loss - potential loss)

• One barrier potential loss and a second barrier potential loss (i.e., potential loss -

potential loss)

At the SITE AREA EMERGENCY classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a GENERAL EMERGENCY is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a GENERAL EMERGENCY classification.

Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the SEC would have greater assurance that escalation to a GENERAL EMERGENCY is less IMMINENT.

CNP Basis Reference(s):

1. NEI 99-01 FS1 Page 173 of 236 INFORMATION USE

ATTACHMENT 1 EAL Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL:

FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-1)

. Mode Applicability:_

1 - Power Operation, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown Definition(s):

None Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers. Table F-1 (Attachment

2) lists the fission product barrier thresholds, bases and references.

At the GENERAL EMERGENCY classification level each barrier is weighted equally. A GENERAL EMERGENCY is therefore appropriate for any combination of the following conditions:

• Loss of Fuel Clad, RCS and Containment barriers

• Loss of Fuel Clad and RCS barriers with potential loss of Containment barrier

• Loss of RCS and Containment barriers with potential loss of Fuel Clad barrier

• Loss of Fuel Clad and Containment barriers with potential loss of RCS barrier CNP Basis Reference(s):

1. NEI 99-01 FG1 INFORMATION USE r Page 174 of 236

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Introduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product

. barrier column is further divided into two columns; one for Loss thresholds and one for.

Potential Loss thresholds.

The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) .of fission product barrier thresholds. The fission product barrier categories are:

A.. RCS or SG Tube Leakage B. Inadequate Heat removal C. CNMT Radiation/ RCS Activity .

D.

• CNMT Integrity or Bypass E. SEC Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories.

The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.

Thresholds are assigned sequential numbers within each Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad barrier Loss in Category A would be assigned "FC Loss A.1," the third Containment barrier Potential Loss in Category C would be assigned "CNMT P-Loss C.3," etc.

If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.

Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers.

When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or pofentialiyfost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the Containment barrier can occur. Barrier Page 175 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Losses and Potential Losses are then applied to the algorithms given in EALs FG1 .1, FS 1.1, and FA 1.1 to determine the appropriate emergency classification.

In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according category Loss followed by category Potential Loss beginning with CategoryA,theri B, ..-., E.

Page 176 of 236 INFORMATION USE I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier Reactor Coolant System (RCS) Barrier Containment (CNMT) Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss

1. An automatic or manual ECCS 1. Operation of a standby charging A

RCS or None None (SI) actuation required by EITHER:

UNISOLABLE RCS .

pump is required by EITHER:

UNISOLABLE RCS leakage 1. A leaking or RUPTURED SG is FAULTED outside of containment None

• SG tube leakage SG Tube Leakage . leakage SG tube RUPTURE

2. CSFST Integrity-RED Path (F-0.4) conditions met

1. CSFST Core Cooling-ORANGE

1. CSFST Core Cooling-RED Path B Path (F-0.2) conditions met 1. CSFST Heat Sink-RED Path (F-0.3) conditions met (F-0.2) conditions met

1. CSFST Core Cooling-RED 2. CSFST Heat Sink-RED Path Inadequate Path (F-0.2) conditions met (F-0.3)conditions met None AND None AND Heat AND Restoration procedures not Heat sink is required Removal effective within 15 min. (Note 1)

Heat sink is required C 1. Containment radiation > Table F-2

1. Containment radiation> Table F-2 1. Containment radiation> Table F-2 CNMT column "FC Loss" None column "RCS Loss" None None column "CNMT Potential Loss" Radiation 2. Dose equivalent 1-131 coolant

/RCS activity > 300 µCi/cc

'Activity

1. Containment isolation is required 1. CSFST Containment-RED Path D .

AND EITHER:

Containment integrity has been lost based on SEC 2.

(F-0.5) conditions met

• Containment hydrogen concentratior 2:4%

CNMT Integrity or Bypass None None None None

. judgment UNISOLABLE pathway from containment to the environment

3. Containment pressure > 2.8 psig with < one full train of depressurization equipment exists operating per design for

2. Indications of RCS leakage 2: 15 min. (Note 1, 9) outside of Containment E 1. Any condition in the opinion of 1. Any condition in the opinion of 1. Any condition in the opinion of 1. Any condition in the opinion of the 1. Any condition in the opinion of 1. Any condition in the opinion of the the SEC that indicates loss of the SEC that indicates potential the SEC that indicates loss of the SEC that indicates potential loss of the SEC that-indicates loss of the SEC that indicates potential loss of SEC the Fuel Clad barrier loss of the Fuel Clad barrier RCS barrier the RCS barrier Containment barrier the Containment barrier Judgment Page 177 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

I None Page 178 of 236 INFORMATION USE---- ,- -

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

None Page 179 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

1. CSFST Core Cooling-REO Path (F-0.2)

. . conditions met Definition(s):

None Basis:*

Indication of continuing severe core cooling degradation is manifested by CSFST Core Cooling RED PATH conditions being met. Specifically, Core Cooling RED PATH conditions exist if either the five highest core exit TCs are reading greater than or equal to 1200°F or core exit TCs are reading greater than or equal to 757°F with RCS subcooling less than or equal 40°F and RVLIS level less than or equal to that specified based on the number of RCPs running (ref. 1) ...

Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery. The CSFSTs are normally monitored using the SPDS display on the PlantProcess Computer (ref 1, 2).

This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant. *

• CNP Basis Reference(s):

1. 1(2)-0HP04023-F-0.2 Critical Safety Function Status Trees...:. Core Cooling

2. 1(2)-0HP-4023-FR-C'.1 Response to Inadequate Core Cooling

3. NEI 99-01 Inadequate Heat Removal F_uel Clad Loss 2.A

,

• Page 180 of 236 INFORMATION USE

• ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

1. CSFST Core Cooling-ORANGE Path (F-0.2) condition$ met Definition(s):

None Basis:-

lndication of continuing significant core cooling degradation is manifested by CSFST Core Cooling ORANGE PATH conditions being met. Specifically, Core Cooling ORANGE PATH conditions exist if either the five highest core exit TCs are reading greater than or equal to 757°F with RCS subcooling less than or equal 40°F or RVLIS level less than or equal to that specified based on the number of RCPs running (ref. 1).

• Critical Safety Function Status Tree (CSFST) Core Cooling-ORANGE path indicates subcooling has been lost and that some fuel dad damage may potentially occur. The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1, 2).

This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat~induced cladding damage.

CNP Basis Reference(s):

1.. 1(2),.QHP04023-F-0.2 Critical Safety Function.Status Trees - Core Cooling

2. 1(2)-0HP-4023-FR-C.2 Response to Degraded Core Cooling

. 3.

• NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A

• *-- j.__*- - - - - ' - - - - - - - - * - *

• _ ** _Pa_g_e_,1_8_1_o_f_23_6_ _ _ _ _ _ _

1N_F_O_R_M_A_T_IO_N~U_S_E____.I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold:

2. CSFST Heat Sink-RED Path (F-0.3) conditions met AND Heat sink is required Definition(s): .

None Basis:

fn combination with RCS Potential Loss B.1, meeting this threshold results in a SITE AREA EMERGENCY.

.. Critical Safety Function .Status Tree (CSFST) Heat Sink-HED path indicates the ultimate heat sink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).

Heat Sink RED PATH conditions exist if narrow range level in all SGs is less than or equal to 13% and total feedwater flow to all SGs is less than or equal to 240,000 lbm/hr (ref. 1).

The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 2).

The phrase lland heat sink required" preclud_es the need for classification for-conditions in which RCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example, FR-H.1 is entered from CSFST Heat Sink-Red. Step 1 tells the operator to determine if heat sink is required by checking that RCS pressure is greater than any non-faulted SG pressure and RCS temperature is greater than 350°F. If these conditions exist, Heat Sink is required. Otherwise, the operator is to either return to the-j:>rocedme and step in effect and place RHR in service for heat removal. For large .

L:QCA E!_Vf:3nts inside th_e:,Containm~nt, the SGs are moot bec:ause heat removal through Jhe containment heat removal systems takes place. Therefore, Heat Sink Red should not be required and, should not be assessed for EAL classification because a LOCA event alone should not require higher than an ALERT classification. (ref. 2).

• 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. In accordance with EOPs, there may be unusual accident 9onditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threstiold is not warranted.

Page 182 of 236

• INFORMATION USE -

I**

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CNP Basis Reference(s):

1. 1(2)-0HP04023-F-0.3 Critical Safety Function Status Trees - Heat Sink

2. 1(2)-0HP-4023-FR-H.1 Response to Loss of Secondary Heat Sink

3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B Page 183 of 236 INFORMATION USE I -~

ATIACHMENT2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CNMT Radiation / RCS Activity Degradation Threat: Loss Threshold:*

1. Containment radiation > Table F-2 column "FC Loss" Table F-2 Containment Radiation - R/hr - VRA-1310 (2310) / 1410 (2410)

• Monitor FC Loss RCS Loss CNMT Potential Loss VRA-1310 (2310) 1,000 200 9,100 VRA-1410 (2410) 700 140 6,300 Definition(s):

None Basis:

Containment radiation monitor readings greater than Table F-2 column "FC Loss" (ref. 1) indicate the release of reactor coolant, with elevated activity indicative of fuel damage, into the containment. The reading is derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 µCi/cc dose equivalent 1-131 into the containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage (2 - 3% clad failure depending on core inventory and RCS volume). This value is higher than that specified for RCS barrier Loss C.1 (ref. 1, 2).

Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors CHRM-VRA-1310/1410 (2310/2410).

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that.expected for iodine spikes and corresponds to an approximate range of 2% to 3% 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 C.1 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 ECL to a SITE AREA EMERGENCY. -

Page 184 of 236 INFORMATION USE I ___ - --- -

- - .I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CNP Basis Reference(s):

1. EP-CALC-CNP-1602, Containment Radiation EAL Threshold Values

2. EVAL-RD-99-11, Evaluation of Radiation Monitoring System Setpoints, Rev 0

3. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.A

-_I Page 185 of 236 INFORMATION USE I_

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CNMT Radiation / RCS Activity Degradation Threat: Loss Threshold:

2. Dose equivalent 1-131 coolant activity> 300 µCi/gm Definition(s):

None Basis:

• This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 3% fuel clad damage (ref. 1).

Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.

There is no Potential Loss threshold associated with RCS Activity/ Containment Radiation.

CNP Basis Reference(s):

1. EP-EALCALC-CNP-1602, Containment Radiation EAL Threshold Values

2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.B Page 186 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CNMT Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

.__N_on_e_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___,I Page 187 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CNMT Integrity or Bypass Degradation Threat: Loss Threshold:

I None Page 188 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CNMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

I No.ne I

Page 189 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. SEC Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the SEC that indicates loss of the Fuel Clad barrier Definition(s):

None Basis:

This threshold addresses any other* factors that* are* to be used* by the Site* Emergency*

Coordinator in determining whether the Fuel Clad barrier is lost CNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A Page 190 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad

• Category: E. SEC Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the SEC that indicates potential loss of the Fuel Clad barrier.

Definition(s):

None Basis:

This threshold addresses any other factors that are to be used by the Emergency SEC in determining whether the Fuel Clad barrier is potentially lost. The SECshould also consider whether or not to declare the barrier potentially l~st in the event that barrier status cannot be monitored.

CNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A

• ~1*_ _ _ _ _ _ _ _ _ _ _ o_f_2_36_ _ _ _ _ _ _*_1N_F_O_R_M_A_I_IO_.N_U_S_E~

• _.P_a_ge_19_1_.

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. An automatic or manual ECCS (SI) actuation required by EITHER:

• UNISOLABLE RCS leakage

• SG tube RUPTURE

• Definition(s):

UNJSOLABLE -Ari open or breached system line that can*not be isolated, remotely or locally.*

RUPTURE - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.

Basis:

ECCS (SI) actuation is caused by (ref. 1, 2):

• Pressurizer low pressure

• Steamline low pressure

• Lower Containment high pressure

• Steamline b.P 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 A.1 will also be met.

CNP Basis Reference(s):

1. 1(2)-0HP-4023-E-O Reactor Trip or Safety Injection

2. 1(2)-0HP-4023-E-3 Steam Generator Tube Rupture

• 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.A Page 192 of 236 INFORMATION USE , .

• I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:*

1. Operation of a standby charging pump is required by EITHER:

• UNISOLABLE RCS leakage .

• SG tube leakage Definition(s):

UNISOLABLE -An open or breached system foie that cannot be isolated, remotely or locally.

• Basis:

This threshold is based on the inability to maintain liquid inventory within the RCS by normal operation of the Chemical and Volume Control System (CVCS). The CVCS includes three charging pumps: one positive displacement pump with a flow capacity of 150 gpm, and two centrifugal charging pumps each with a flow capacity of 150 gpm (ref. 1). A second charging pump being required is indicative of a substantial RCS leak.

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 EGGS (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 1.A will also be met.

CNP Basis Reference(s):

1. UFSAR Table 9.2-2 Chemical and Volume Control System Design Parameters .

2. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.A Page 193 of 236 INFORMATION USE 1*

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

2. CSFST Integrity-RED Path (F-0.4) conditions met Definition(s):

None Basis:

• The "Potential Loss" threshold is defined by the CSFST Integrity - RED path. CSFST Integrity -

Red Path plant conditions and *associated PTS Limit Curve A indicates an extreme challenge to the safety function when plant parameters are to the left of the limit curve following excessive RCS cooldown under pressure (ref. 1, 2).

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).

CNP Basis Reference(s):

1. 1(2)-0HP-4023-F-0.4 Critical Safety Function Status Trees Figure F-0.4-1 Integrity Operational Limits

2. 1(2)-0HP-4023-FR-P.1 Response to Imminent Pressurized Thermal Shock Condition

3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.8 Page 194 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

I None Page 195 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Potential Loss Threshold: *

1. CSFST Heat Sink-RED Path (F-0.3) conditions met ANb Heat sink is required Definition(s):

• None*

Basis:

In combination with FC Potential Loss B.2, meeting this threshold results in a SITE AREA EMERGENCY .

. Critical Safety Function Status Tree (CSFST) Heat Sink-RED path indicates the ultimate heat sink function is under extreme challenge and that some fuel clad .damage may potentially occur (ref. 1).

• Heat Sink RED PATH conditions exist if narrow range level in all SGs is less than or equal to 13% and total feedwater flow to all SGs is less than or equal to 240,000 lbm/hr (ref. 1).

The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 2).

The phrase "and heat sink required" precludes the need for. classification for conditions in

. which RCS pressure is less than SG pressure or Heat Sink-RED path entry was created through operator action directed by an EOP. For example, FR-H.1 is entered from CSFST Heat Sink-Red. Step 1 tells the operator to determine if heat sink is required by checking that RCS pressure is greater than ariy non-faulted SG pressure and RCS temperature is greater than 350°F. If these conditions exist, Heat Sink is required. Otherwise, the operator is to either

-* return to the procedure and step in effect and place RHR in service for h*eat removal. For large LOCA events j11side the Conta_inm~_nt, the SGs are moot because heat removal through the containment heat removal systems takes place. Th_erefore, Heat Sink Red should not be required and, should not be assessed for EAL classification because a LOCA event alone should not require higher than an ALERT classification. (ref. 2).

This condition indicates an extreme-challenge to the ability to remove RCS heat using the steam generators (i.e., loss ofan effective secondary-side--heat-sink). This condition*

. represents 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 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 B.2; both will be met. This condition

-warrants a SITE AREA EMERGENCY declaration becam~e inadequate RCS heat-removal may

• • !* *Page 196 of 236 * *. INFORMATION USE* /. **

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 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.

CNP Basis Reference(s):

1. 1(2)-0HP04023-F-0.3 Critical Safety Function Status Trees - Heat Sink

2. 1(2)-0HP-4023-FR-H.1 Response to Loss of Secondary Heat Sink

3. NEI 99-01 Inadequate Heat Removal RCS Loss 2.8 Page 197 of 236 INFORMATION USE

. I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: C. CNMT Radiation/ RCS Activity Degradation Threat: Loss Threshold:

1. Containment radiation > Table F-2 column "RCS Loss" Table F-2 Containment Radiation - R/hr - VRA-1310 (2310) / 1410 (2410)

• Monitor

• FC Loss RCS Loss CNMT Potential Loss VRA-1310 (2310) 1,000 200 9,100 VRA-1410 (2410) 700 140 6,300 Definition(s):

N/A Basis:

Containment radiation monitor readings greater than Table F-2 column "RCS Loss" (ref. 1, 2) indicate the release of reactor coolant to the containment. The readings assume the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the containment atmosphere. Because of the very high fuel clad integrity, only small amounts of noble gases would be dissolved in the primary coolant.

Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors CHRM-VRA-1310/1410 (2310/2410).

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 C.1 since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with RCS Activity/ Containment Radiation.

CNP Basis Reference(s):

1. EP-CALC-CNP-1602, Containment Radiation EAL Threshold Values

2. EVAL-RD-99-11, Evaluation of Radiation Monitoring System Setpoints, Rev 0

3. NEI 99-01 CMT Radiation/ RCS Activity RCS Loss 3.A Page 198 of 236 INFORMATION USE I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. CNMT Radiation/ RCS Activity Degradation Threat: Potential Loss

.Threshold:

None I

Page 199 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CNMT Integrity or Bypass Degradation Threat: Loss Threshold:

I None Page 200 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CNMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

I None

, '~-----------Pa_g_e_2_0_1_o_f_23_6_ _ _ _ _ _ _ 1N_F_O_R_M_A_T_IO_N_U_S_E__,

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. SEC Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the SEC that indicates loss of the RCS barrier Definition(s):

None Basis:*

This threshold addresses any other factors that may be used by the Site Emergency Coordinator in determining whether the RCS Barrier is lost.

CNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A Page 202 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. SEC Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the SEC that indicates potential loss of the RCS barrier Definition(s):

None

-Basis:

This threshold addresses any other factors that may be used by the Site Emergency Coordinator in determining whether the RCS Barrier is potentially lost. Th*e Site Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

CNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A Page 203 of 236 INFORMATION USE I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. RCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. A leaking or RUPTURED SG is FAULTED outside of containment Definition(s):

• FAUL TED - The term applied to a steam generator that has a steam leak on the secondary side .of sufficient size to cause an uncontroll~d drop in steam generator pressure or the steam generator to become completely depressurized.

RUPTURED - The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.

Basis:

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 A.1 and Loss A.1, 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~ 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 FAULTED for emergency classification purposes.

The FAULTED 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 SUS 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 power operated relief valve or safety relief valve do not meet the intent of this threshold. Such release.s 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 I Page 204 of 236 INFORMATION USE 1*

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.

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 loss of containment but should be evaluated using the Recognition CategoryR ICs.

The EC Ls resulting from primaiy-to~secoridary 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 UNUSUAL EVENT per UNUSUAL EVENT per Greater than 25 gpm SU5.1 SU5.1 Requires operation of a standby SITE AREA charging (makeup) pump (RCS ALERT per FA 1.1 EMERGENCY per FS1 .1 Barrier Potential Loss)

Requires an automatic or manual SITE AREA ECCS (SI) actuation (RCS Barrier ALERT per FA1.1 EMERGENCY per FS1 .1 Loss)

There is no Potential Loss threshold associated with RCS or SG Tube Leakage.

CNP Basis Reference(s):

1. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.A Page 205 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. RCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

Page 206 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. Inadequate heat Removal Degradation Threat: Loss Threshold:

I None Page 207 of 236 INFORMATION USE

ATTACHMENT 2

• Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: B. Inadequate heat Removal Degradation Threat: Potential Loss Threshold:

1. CSFST Core Cooling-RED Path (F-0.2) conditions met AND ReistoraUon procedures not effective within 15.min. (Note 1)

. Note 1: . The SEC should declare the event promptly upon determining that time limit has been exceeded, or will .

likely be exceeded:

• Definition{s):

None Basis:

Indication of continuing severe core cooling degradation is manifested by CSFST Core Cooling

• REO PATH conditions. being met. Specifically, Core Cooling RBI]) PATH conditions exist if either the five highest core exit TCs are reading greater than or equal to 1200°F or core exit TCs are reading greater than or equal to 757°F with RCS subcooling less than or equal 40°F and RVLIS level less than or equal to that specified based on the number of RCPs running (ref. 1).

Critical Safety Function Status Tree.(CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery. The .CSFSTs are normally monitored using the SPDS display on 'the Plant Computer (ref. 1).

The function restqration procedures are those emergency operating procedures that address the.recovery of the core cooling critical safety-functions. The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing (ref. 1, 2, 3).

A direct correlation to status trees can be made if.the effectiveness of the restoration procedures .

__ i~.also evaluatec:!. If core exitthermocouple (TC) re~c:!Jngs are greater than 1,20.0~F (ref. :I), Fu.el ...

Clad barrier is also lost.

This threshold addresses any other factors that may be used by the Site Emergency

• Coordinator in determining whether the RCS Barrier is potentially lost. 1he Site Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the

-

• event-that barrier status cannot be monitored:

• cN*P Basis Referen~e{s):.

1. 1(2)-0HP04023-F-0.2 Critical Safety Function Status Trees - Core Cooling

2. 1(2)-0HP-4023-FR-C.1 Response to Inadequate Core Cooling *

3. 1(2)-0HP04023-FR-C.2 Response to Degraded Core Cooling

• - 4. NEI 99-01 Inadequate Heat Removal Containment Potential Loss. 2.A *

-

• Page 208 of 236 * -* -- - INFORMATION* USE_. , _

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CNMT Radiation/RCS Activity Degradation Threat: Loss Threshold:

/'------Non*--------'---------'

Page 209 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CNMT Radiation/RCS Activity Degradation Threat: Potential Loss Threshold:

1. Containment radiation > Table F-2 column "CNMT Potential Loss" Table F-2 Containment Radiation - R/hr - VRA-1310 (2310) / 1410 (2410)

Monitor

• FC Loss

• RCS Loss CNMT Potential Loss

• VRA-1310 (2310) 1,000 200 9,100 VRA-1410 (2410) 700 140 6,300 Definition(s):

None Basis:

Containment radiation monitor readings greater than Table F-2 column "CNMT Potential Loss" (ref. 1, 2) indicate significant fuel damage {20% clad damage) well in excess of that required for loss of the RCS barrier and the Fuel Clad barrier.

The readings are higher than that specified for Fuel Clad barrier Loss C.1 and RCS barrier Loss C.1. Containment radiation readings at or above the containment barrier Potential Loss threshold, therefore, signify a loss of two fission product barriers and Potential Loss of a third, indicating the need to upgrade the emergency classification to a GENERAL EMERGENCY.

Monitors used for this fission product barrier loss threshold are the Containment High Range Radiation Monitors CHRM-VRA-1310/1410 (2310/2410).

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 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 of radioactivity 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 ECL to a GENERAL EMERGENCY.

Page 210 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CNP Basis Reference(s):

1. EP-CALC-CNP-1602, Containment Radiation EAL Threshold Values

2. EVAL-RD-99-11, Evaluation of Radiation Monitoring System Setpoints, Rev 0

3. NEI 99-01 GMT Radiation / RCS Activity Containment Potential Loss 3.A Page 211 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CNMT Integrity or Bypass Degradation Threat: Loss Threshold:

1. Containment isolation is required AND EITHER:

• Containment integrity has been lost based *on SEC judgment

• UNISOLABLE. pathway from containment to the environment exists Definition(s):

UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally.

Basis:

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A.1.

These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. Users are reminded that there may be accident and release conditions that simultaneously meet both bulleted thresholds.

First Threshold - 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). 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 Site Emergency Coordinator 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 1. 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.

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 I Page 212 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases or potential loss of containment but should be evaluated using the Recognition Category R ICs.

Second Threshold - 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 conta111ment

• pressure.

Refer to the top piping run of Figure 1. 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 existencef of a filter is not considered in the threshold assessment. *Filters do not reniove

• 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.

Leakage between two interfacing liquid systems, by itself, does not meet this threshold.

Refer to the bottom piping run of Figure 1. 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 developed a leak that allowed steam/water to enter the Auxiliary Building, then second threshold 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 the first threshold to be met as well.

Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable 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 an enclosed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.

CNP Basis Reference(s):

1. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.A Page 213 of 236 INFORMATION USE

.I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CNMT Integrity or Bypass Degradation Threat: Loss

. threshoid: .

2. Indications of RCS leakage outside of containment Definition(s):

None Basis:

• The status of the containment barrier during an event involving steam generator tube leakage

• is assessed using Loss Threshold A.1.

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 A.1 to be met.

ECA-1.2 LOCA Outside Containment (ref. 1) provides instructions to identify and isolate a LOCA outside of the containment. Potential RCS leak pathways outside containment include (ref. 1, 2):

• Residual Heat Removal

• Safety Injection

• Chemical & Volume Control

• RCP seals 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 1. 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 D.1 to be met as well.

Page 214 of 236 INFORMATION USE 1*

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CNP Basis Reference(s):

1. 1(2)-0HP-4023-ECA-1.2 LOCA Outside Containment

2. 1(2)-0HP-4023-E-1 Loss of Reactor or Secondary Coolant

3. NEI 99-01 CMT Integrity or Bypass Containment Loss Page 215 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Figure 1: Containment Integrity or Bypass Examples Auxiliary Building Inside Reactor Building Damper RCP Seal Cooling Page 216 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CNMT Integrity or Bypass Degradation Threat: Potential Loss Threshold: --

1. CSFST Containment-RED Path (F-0.5) conditions met Definition(s):

None

-Basis:

Critical Safety Function Status Tree (CSFST) Containment-RED path is entered if containment pressure is greater than or equal to 12 psig and represents an extreme challenge to safety function. The CSFSTs are normally monitored using the SPDS display on the Plant Computer (ref. 1, 2).

12 psig is the containment design pressure (ref. 3) and is the pressure used to define CSFST Containment Red Path conditions.

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.

CNP Basis Reference(s):

1. 1(2)-0HP-4032-F00.5 Critical Safety Function Status Trees Containment

2. 1(2)-0HP-4023-FR-Z.1 Response to High Containment Pressure

3. UFSAR Section 5.2.2.2 Design Load Criteria

4. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.A Page 217 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CNMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

2. Containment hydrogen concentration ;;:: 4%

Definition(s):

None Basis:

Following a design basis accident, hydrogen gas may be generated inside the containment by reactions such as zirconium metal with water, corrosion of materials of construction and radiolysis of aqueous solution in the core and sump. The lower limit of combustion of hydrogen in air is approximately 4%.

CNP is equipped with a Post-Accident Hydrogen Monitoring System (PACHMS) which serves to measure combustible gas concentrations in the containment. The PACHMS is comprised of two sampling-analyzing-control trains (ref. 1).

To generate such levels of combustible gas, loss of the Fuel Clad and RCS barriers must have occurred. With the Potential Loss of the containment barrier, the threshold hydrogen concentration, therefore, will likely warrant declaration of a GENERAL EMERGENCY.

Two Containment hydrogen monitors with duel ranges of 0% to 10% and 0% to 30% provide indication locally and in the Control Room (ref. 1, 2).

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.

CNP Basis Reference(s):

1. UFSAR Section 7.8.2 Post-Accident Hydrogen Monitoring

2. 12-THP-6020-PAS-003, Post Accident Containment Hydrogen Monitoring System Operation

3. NEI 99-01 GMT Integrity or Bypass Containment Potential Loss 4.8 Page 218 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CNMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

3. Containment pressure > 2.8 psig with < one full train of containment depressurization equipment operating per design for;;:: 15 min. (Notes 1, 9)

• Note 1: The SEC should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 9: One Containment Spray System train and one Containment Air Recirculation Fan comprise one full

.train .of depress.urization equipment.

Definition(s):

None Basis:

Containment pressure control is achieved through the Containment Spray System and the Containment Air Recirculation/Hydrogen Skimmer System. Failure of either of these systems may allow steam to build up within containment, and, unabated, this steam buildup may cause the internal containment pressure buildup to exceed the design pressure of 12 psig. Studies have shown that the containment can withstand pressures well above this value.

Both the recirculation fans and the containment spray pumps are actuated automatically (delayed) following receipt of a HI or HI HI containment pressure signal, respectively. Since the HI HI containment pressure setpoint is less than or equal to 2.8 PSI, then greater than 2.8 PSI would be the containment pressure greater than the setpoint at which the equipment was supposed to have actuated. If these systems should fail to start automatically per design, a successful manual start within 15 minutes would preclude exceeding this Containment Potential Loss threshold. (ref. 1, 2, 3).

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 recirculation fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.

Page 219 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CNP Basis Reference(s):

1. UFSAR Section 5.5.3 System Description

2. UFSAR Section 6.3 Containment Spray Systems

3. EC-0000052930 Unit 1 Return to Normal Operating Pressure and Temperature (NOP/NOT)

4. NEI 99~01 GMT Integrity or Bypass Containment Potential Loss 4.C Page 220 of 236 INFORMATION USE

. I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: E. SEC Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the SEC that indicates loss of the Containment barrier Definition(s):

None Basis:

This threshold_ addresses any other facto_rs th_at may be used by the Site Emergency _

Coordinator in determining whether the Containment Barrier is lost.

CNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A Page 221 of 236 INFORMATION USE

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: E. SEC Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the SEC that indicates potential loss of the Containment barrier Definition(s):

None Basis:

This threshold addresses any other factors that may be used by the Site Emergency Coordinator in determining whether the Containment Barrier is lost.

CNP Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A Page 222 of 236 INFORMATION USE

ATIACHMENT3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases

Background

NEI 99-:-01 Revision 6 ICs AA3 and HA5 prescribe declaratio.n of an ALERT based on IMPEDED access to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located. These areas are intended to be plant operating mode dependent.

• Specifically the Developers Notes For AA3 and HA5 states:

The "site-specific list of plant rooms or areas with entry-related 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 plant operation, coo/down 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 address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.

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).

Further, as specified in IC HA5:

The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous 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 positive pressure within the Control Room envelope.

. . _ I_ _ _ _ _ _ _ _ _ _ _ _ P_ag_e_2_2_3_o_f_23_6_ _ _ _ _ _ _ 1N_F_O_R_M_A_T_IO_N_U_S_E~

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases CNP Table R-2 and H-2 Bases NEI 99-01 Revision (Rev.) 6 addresses elevated radiation levels or hazardous gases 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 cool down and shutdown.

1(2)-0HP-4021-011-001, At-Power Operation Including Load Swings (Rev. 37(36)), was reviewed to determine if any actions are "necessary" to maintain power operations. Over reasonable*

periods of time (days vice months or years) there are no actions outside the Control Room that are required to be performed to maintain normal operations. Eventually, you would have to shut down if Technical Specification (T.S.) surveillance testing was not completed and you

• complied with the associated Limiting Condition for Operations based on consumable supplies being depleted.

The only Emergency Operating Procedures (EOP) reviewed were those that are normally trarisitiohed through during all shutdowns at Cook (1 (2)-0HP-A023-E-O, Reactor Trip or Safety*

Injection, and 1(2)-0HP-4023-ES-0-1, Reactor Trip Response)

The only Abnormal Operating Procedure (AOP) reviewed was 1(2)-0HP-4022-001-006, Rapid Power Reduction, as it could have also been used during the initiation of a plant shutdown.

The following table lists the locations into which an operator (or Chemistry Technician) may be dispatched in order to perform a normal plant cool down and shutdown. Chemistry Technician was included due to the fact that obtaining and analyzing for the Reactor Coolant System (RCS) boron concentration is an integral action in order to achieve and maintain cold shutdown (CSD) conditions. The review was completed using the following procedures as the controlling documents:

• 1(2)-0HP-4021-011-001, At-Power Operation Including Load Swings (Rev. 37(36))

• 1(2)-0HP-4021-001-003, Power Reduction (Rev. 60(55))

• 1(2)-0HP-4022-001-006, Rapid Power Reduction (Rev. 16(15))

• 1(2)-0HP-4023-E-O, Reactor Trip Or Safety Injection (Rev. 41(41))

• 1(2)-0HP-4023-ES-0-1, Reactor Trip Response (Rev. 30(29))

• 1(2)-0HP-4021-001-004, Plant Cooldown From Hot Standby To Cold Shutdown (Rev. 78(69))

• 1(2)-0HP-4021- 017-002, Placing In Service The Residual Heat Removal System (Rev. 31(28))

Each step in the controlling procedures was evaluated to determine if the action was performed in the Control Room or in the plant. Each in-plant action listed below was evaluated and a determination made whether or not the actions, if not performed, would prevent achieving CSD. The following generic assumptions were applied:

• Steps involving optional degasing of the RCS were not selected since degasing the RCS is not required to reach cold shutdown.

• Steps involving Main Feed Water Pumps were not selected since Auxiliary Feedwater (AFW) can be used to reach cold shutdown if Main Feed Water is not available.

• Steps that are stated as needed when entering an outage are disregarded, as they are optional and not mandatory for placing plant in CSD.

Page 224 of 236 INFORMATION USE

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Travel paths to the locations where the equipment is operated are not part of the determination of affected room/area, only the rooms/areas where the equipment is actually operated. Most locations can be reached via alternate travel paths if required due to a localized issue.

No assumption is made about which Residual Heat Removal (RHR) Train is aligned for operation.

The minimum set of in-plant actions, associated locations, and operating modes to shut down arid cool down the reactor are highlighted. The locations where those actions are performed comprise the rooms/areas to be included in Emergency Action Level Tables R-2 and H-2.

CRITERION.11 Control Room (From CNP UFSAR)

The facility shall be provided with a control room from which actions to maintain safe operational status of the plant can be controlled. Adequate radiation protection shall be provided to permit continuous occupancy of the control room under any credible post-accident

• condition or as an alternative, access to other areas of the facility as necessary to shutdown .

and maintain safe control of the facility without excessive radiation exposures of personnel.

Each unit of the plant is equipped with a separate control room, which contains those controls and instrumentation necessary for operation of that unit under normal, and accident conditions.

The control room is continuously occupied by the operating personnel under all operating and accident conditions, unless the control room should become uninhabitable. This case is discussed in Section 7.7.10.

Sufficient shielding, distance, and containment integrity are provided to assure that control room personnel shall not be subject to doses under postulated accident conditions during occupancy of the control room which would exceed 10% of the limits suggested in 10 CFR 100. The control room ventilation system is discussed in Chapter 9 of the UFSAR.

• • Building/* a, If action not performed, Procedure and

'Step Actio,n , .,

• Elevation/* '"g does, this. prevent 'GOO!

Step

,, , .,. , * ,. . Ro9m., . , ~ c down/ ,shyt doiivp.?.

'*., ',>'>,,,. '>>>*,, ,'f*'  ;' ,;;, ' *L d ,, ;;,.,' **.'  :* . *: , :,/.,,,;,;'. c Attachment (Att.) When reducing power monitor Auxiliary (Aux.) No - Blowdown could be 2, Step 3 .10 (1 & blowdown flash tank pressure (local 591' or 633' isolated from the Control 1

2) indication only) depending on Room. Not monitoring would tank in service not prevent achieving CSD.

Att. 2, Step 4.5 (1 Maintaining Main Generator Turbine (Turb.) No - Not monitoring Main

&2) parameters 633', 609', 591' Generator parameters would Att. 3, Step 4.6 (1 not prevent achieving CSD.

&2)

Att. 3, Step 4.4 & Maintaining Volume Control Tank Aux. 609' No - Not monitoring VCT 4.12 (1) (VCT) pressure pressure would not prevent achieving CSD.

Both units per Placing in service and removing the Aux. 609' No - Not maintaining RCS Chemistry request Cation Demineralizer to control pH causes chemistry concerns (usually daily) lithium concentration (pH controi) but does riot prevent ' '

achieving CSD.

Page 225 of 236 INFORMATION USE

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases

_1(2)-0HP-4021-001-003, Power Reduction (Rev. 60(55))

Step 4.3 Boration/Rapid Boration, Lineups to Aux. 587', 573', >:ve1 ~:R~iHgiiine~t o{ihe <*

Step 4.6.1.b. ensure sufficient boric acid volumes 609' B.o~i6 A~icl System arid *

. poterifoil makeup from the" Boric Acid Reserve Tank (BART) is.the basis bf*::

Jocation. and the abilitfto}.

3i41 h::fill.tlie:SoripAcid Storage.**,

5 *Tani<(BAST) iii ordeltc;, :havf .*

• sufficient boric acid to reach" CSD Boron. concentrati~n>

BART. would only be*

.* required}cir .dual unit .* ..

• sh~tdpwn to C~D (Au)c .

"573'_)".-_. . ...

Step 4.9.4 (1) Align all available Steam Jet Air Turb. 609' No - Needed to maintain Step 4.10.4 (2) Ejector (SJAE) Elements condenser vacuum however 112/ cooldown can be achieved 314 with Steam Generator (SIG)

Power Operated Relief Valves (PORV).

Step 4.16 (1) Verify Auxiliary Steam is supplied U1 -Turb. 591' No - Steam Seals and SJAEs from the opposite unit. 1/2/ not required to utilize SIG Step 4.20 (2) U2 - Turb. 609', 314 591' PORVs.

Step 4.25.1.j (1) Upon opening the Reactor Trip NIA No - 1(2)-0HP-4023-E-O, Step 4.28.12 (2) Breakers at between 15 -17% power, Reactor Trip or Safety 1

transition to 1(2)-0HP-4023-E-O, Injection will be reviewed Reactor Trip or Safety Injection. separately in this document.

Step 4.30 (1) If unit is not to be removed from Turb. 591 ', Low Flow Feed water Step 4.32 (2) service at between 15 - 17% power 609', 633' Preheating not necessary for and a Main Feed Pump (MFP) is to the unit to get to CSD.

1 remain in service then place Low Flow Feedwater Preheating in service per OHP-4021-055-002.

Step 4.33.1.i (1) Upon opening the Reactor Trip NIA No - 1(2)-0HP-4023-E-O, Breakers at power levels< 15%, Reactor Trip or Safety transition to 1 Injection, will be reviewed 1(2)-0HP-4023-E-O, Reactor Trip or separately in this document.

Safety Injection.

Step 4.59 (1) Initiate Containment Leak Inspection Containment No - Entry into Containment Step 4.58 (2) per 1-0HP-4021-001-002, Reactor to complete the leak 3/4 Start-Up inspection will not prevent achieving CSD.

Step 4.61.2 (1) Verify the CRDM-MG sets are Aux. 609' No - Technical Specifications Step 4.60 (2) removed per 1(2)-4021-012-001, for Control Rods not being 3/4 Operation of the Control Rod Drive capable of withdrawal can be System.

Page 226 of 236 INFORMATION USE I

i

-1

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases verified from the Control Room.

Step 4.68 (1) Close the Main Generator Hydrogen Turb. 609' No - Not required to achieve Step 4.67 (2) Cooler Turbine Auxiliary Cooling 3/4 CSD.

Water Outlet Valves.

Step 4.71 (1) Removing Main Transformer Cooling Outside 609' No - Not required to achieve 3/4 CSD.

Step 4.70 (2)

Step 4.71 (2) Dispatch an Operator to verify that Outside 609' No - Not required to achieve the Main Transformer Cooling was CSD.

3/4

• automatically removed from service when tlie Unit was tripped. (U2 only)

HR~~~~,~l~~nQ~,1~[iel~1!i~~~;:~ffstt:~

Step 12.a, 2nd Monitor and adjust Stator Water and Turb. 591', 609' No Bullet (1) Generator Hydrogen Temperature 1

Step 13 .a, 2nd Bullet (2)

Step 12.a, 3rd Monitor and adjust Steam Generator Aux. 591' or No - S/G blowdown can be Bullet (1) blowdown flash tank pressure 633' (depending removed from service from on in service 1 the Control Room if Step 13 .a, 3rd Bullet (2) Flash Tank) necessary.

Step 12.d (1) Initiate transfer of Aux Steam as Turb. 591', No - Plant can be cooled Step 13.d (2) required: Refer to 12-0HP-4021-061- Turb. 609'(Unit down utilizing atmospheric 002, Operation of Auxiliary Steam 2 only) 1/2/ steam dumps System, this would support Main 3/4 Turbine steam seals in order to maintain Main Condenser vacuum.

Step 20.b RNO (1) Transition to 1(2)-0HP-4023-E-O, NIA No - 1(2)-0HP-4023-E-O, Step 21.b RNO (2) Reactor Trip or Safety Injection if< Reactor Trip or Safety 17% power and staffing is not 1 Injection. will be reviewed adequate for maintaining Steam separately in this document.

Generator level control in manual.

Step 21.c (1) Starting Auxiliary Feed Water (AFW) Turb. 591' No - AFW is in standby Step 22.c (2) Pumps per 1(2)-0HP-4021-056-002, readiness for automatic start establishing minimum flow and long term cooling and/or 1/2/

requirements. minimum flow requirement 3/4 are not necessary for a necessary for a CSD required shutdown.

Step 23.b (1) Chemistry sample RCS and Aux. 587', 609' No - RCS boron sample for Step 24.b (2) radioactive gaseous effluents intent of (Nuclear iodine and entry into an event this step is to sample for the iodine Sampling Room, initiated surveillance would 1/2/

spike. Chemistry Hot not impact achieving CSD.

3/4 Lab, Alternate

• sample point in

• RHRHxRoom)

Page 227 of 236 INFORMATION USE

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases 1(2)-0HP-4023-E-O, Reactor Trip or NIA At Cook the normal Safety Injection normal Reactor Trip procedural flow path in the

. response immediate actions. applicable modes when performing a reactor trip is to enter 1(2)-0HP-4023-E-O, Reactor Trip or Safety Injection. There are no in 1/2/

plant action required upon a 3

normal plant trip. At step 4, on a routine plant trip response, there is a transition to 1(2)-0HP-4023-ES~O-l, Reactor Trip Response, which is reviewed separately in this document.

1(2)-0HP-4023-ES-O-l, Reactor Trip Response, is in the normal procedural flow path, in the applicable modes, when responding to a normal reactor trip. Expected transition is then to 1(2)-

0HP-4021-001-004, Plant Cooldown From Hot Standby To Cold Shutdown.

Step Lb RNO 5.c Locally close any known open steam Aux. 612' (Feed No - Valves being left open line warming valves. Reg. Valve would not prevent achieving 3/4

• 1-MS-143 / 2-MS-147 Area) CSD.

• 1-MS-144 / 2-MS-148 Step 12 Transfer Aux Steam Supply Turb. 591' (1) No - Plant can be cooled Turb. 591 ', 609' 3/4 down utilizing atmospheric (2) steam dumps.

Step 15.b Shut down the Turbine Driven Turb. 591' No - Shutdown TDAFP and Auxiliary Feedpump (TDAFP) and (Aux. 591' and placing in standby not place in standby (if both Motor 609' U2 only) necessary to continue plant Driven Aux. Feedwater Pumps cooldown to CSD if both (MDAFP) running). Refer to MDAFPs are running. The 1-0HP-4021-056-002, Auxiliary Ul only requirement Feedpump Operation Att.l. 3 requiring access to Aux. 591' and 609' is for placing validating locally the proper positioning of the TDAFP motor operated discharge valves to each Steam Generator .

. . _ I_ _ _ _ _ _ _ _ _ _ _ _ P_a_g_e_2_2_a_o_f_23_6_ _ _ _ _ _ _1N_F_O_R_M_A_T_1o_N_u_s_E---'

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases If afa~on 'nqttt~rfor'*ed, o does thi.s pr,e"~nt ce>ol

~ t, do.W9J,~~u~J~§v.r9{~*};;t* .

Step 16.b This is the transition step to go to the NIA No -This is the expected appropriate plant procedure, which transition for the plant to go would be 1(2)- OHP-4021- 001-004, to a CSD condition post Plant Cooldown From Hot Standby reactor trip.

To Cold Shutdown.

3 t(2}-0HPAC>21 :001-004, Plant Cool,doWn FrQm Hot"Staodby To :cold ShutdQWn *(~ev.. 78(~9))

' ',,' {.,' , ' ' ,, ,l,, j ~,- ~ ' - .' >.*",

Step 4.1.1 1'1 Option to borate the RCS per 1(2)- Aux. 573', 587', Ye~ Iieali~nin~nt 6fih~

Bullet 0HP-4021-005-002, Operation of the 609' (BART Boric' Acid :s*ystem and Unit 1 Boric Acid Blender Room, Boric '.potential ,rriakiup from tlie : *.* .

Acid Storage

• BARr*isJhe h~sis qflocation * .

Tank Room, . arid the ability to refillth~

Borid Acid 3/4/

• BASt in orpe(to have*.*.

Batch Tank 5 sufficient ~ori,c .acid to*re11ch .

Area) CSD Boron concentration.

• B~T )YOuld orily be. * .

required for dual'unit . .

shut<lpwn to CSD (Aux.'. *.

• 573?).: . . ..

Step 4.1.1 3rd Emergency Boration per 1(2)-0HP- Aux. 573', 587', Ye~ ~ Realignment or:the Bullet 4021-005-007, Operation of 609' (BART .Boric Apid System and .

Emergency Boration Flow Paths. Room, Boric >pqtential mak~up frc;np)he '

Acid Storage BART is the basis of!ocatibn Tank Room, and th¢ ability to. refill tJ:ie BoridAcid 3/4/ BASTin order to havi:"" * .

Batch Tank 5 suffi~ie~t b~fic add t~ ;~ach .

Area) . CSD Boron concentration ..

~,i\R'f wo.ui<lNily 6-e :* . *. *

>required:for d~al unit: ....... .

shutdow:n to CSD (Aux.-

5}3'.j.:: . . .

Att. 5 Step 4.17.1 Rack Out Reactor Trip and Bypass Aux. 609' (4KV No - Racking out Reactor Att. 6 Step 4.18.1 Breakers Room) Trip and Bypass Breakers 3/4 does not impact achieving Att. 7 Step 4.18.1 CSD.

Att. 8 Step 4.20.1 Att. 5, Cooldown U1 - Initiate l-OHP-4030-066-4025, Various No- Validating opposite unit Using the Unit 1 NFPA 805 and Ventilation shutdown equipment support Condenser Steam Requirements for Unit 2. would not prevent achieving Dumps, Step 4.18 U2 - Initiate 2-0HP-4030-066-4025, 4 CSD.

Unit 2 NFPA 805 and Ventilation Att. 6, Cooldown Requirements for Unit 1.

using the Steam **********************n****

Page 229 of 236 INFORMATION USE

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Generator Power Operated Relief Valves, Step 4.19 Att. 7, Plant Cooldown using Condenser Steam Dumps with Solid Plant Operations, Step 4.19 Att. 8, Plant Cooldown using Steam Generator PORV's with Solid Plant Operations, Step 4.21 Att. 5 Step 4.29.2 Maintenance Instrumentation (MTI) Containment No - Not required to be (U2, Step 4.29.1) perform performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after Att. 6 Step 4.30.2

• 1(2)-IHP-4030-102(202)-01 OA, decreasing RCS cold leg Train A Pressurizer PORV temperature to~ 266°F.

Att. 7 Step 4.30.2 Actuation Channel Functional Performed to satisfy T.S.

Att. 8 Step 4.32.2 3/4 3.4.12.8. Therefore not Test (LTOP)

• 1(2)-IHP-4030-102(202)-01 OB, completing this step would Train B Pressurizer PORV not prevent achieving CSD.

Actuation Channel Functional Test (LTOP)

Att. 5 Step 4.29.3 Maintenance Instrumentation (MTI) Containment No - Performed to satisfy T.S.

(U2, Step 4.29.2) perform 8.4.6.1. Not completing this Att. 6 Step 4.30.3

• 1(2)-IHP-4030-102(202)-0llA, step would not prevent Train A Pressurizer PORV 1- achieving CSD.

Att. 7 Step 4.30.3 NRV-153 Emergency Air System Att. 8 Step 4.32.3 Channel Functional Test and 3/4 Calibration

• 1(2)-IHP-4030-102(202)-01 lB, Train B Pressurizer PORV 1-NRV-152 Emergency Air System Channel Functional Test and Calibration Att. 5 Step 4.29.4 Operation perform applicable sections Containment No - Only required if (U2, Step 4.29.3) of 1(2)-0HP-4030-102(202)-060, adjustments are made to the Att. 6 Step 4.30.4 PRZ Power Operated Relief Valve PORV(s) in step 4.29.2, Testing, If adjustments were made 3/4 which is not required to be Att. 7 Step 4.30.4 during 1(2)-IHP-4030-102(202)-0lOA performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after Att. 8 Step 4.32.4 or 1(2)-IHP-4030-102(202)-0lOB. decreasing RCS cold leg temperature to~ 266°F.

Therefore not completing this Page 230 of 236 INFORMATION USE

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases

• . Building'l{'}

Elevation/ i

. *./i'iilR~ci61.\.

step would not prevent achieving CSD.

Att. 5 Step 4.33.1 When RCS pressure is < 1000 psig Aux. 609 (4KV y ~~}basis is th~t without Att. 6 Step 4.33.l then close, de-energize, and establish Room,* closing Accumulator outlet .

Administrative LTOP controls on Mezzanine , valves, RCS pressm:e can.bot.**.**

Att. 7 Step 4.34.1 'go:beiow ::;6~0 psig./(si ' .. ' '

Accumulator Outlet Valves. Area)

Att. 8 Step 4.35.1 3/4 3.5.1) * ,, .

• EZC-C-5C for IM0-110

• EZC-B-lC for IM0-120

• EZC-D-lC for IM0-130

• EZC-A-5C for IM0-140 Att. 5 Step 4.33.7 Initiate performance of 1(2)-0HP- Au*x. 609 (4KV No~ Not performing the Att. 6 Step 4.33.7 4030-117-054I, RHR Suction Valve Room, applicable testing does not Interlock Test. Mezzanine 3/4 prevent achieving CSD.

Att. 7 Step 4.34.7 Area)

Att. 8 step 4.35.7 Att. 5 Step Establishing LTOP controls for the Aux. 587' No -There are other means 4.35.1.e Safety Injection System. Option to of establishing the SI System (U2, Step Close L TOP conditions.

4.35.1.e)

• SI-11 lN and Att. 6 Step

• SI-111 S, SI Pump Discharge 4.35.1.e Valves.

3/4 Att. 7 Step ******************************

4.37.1.e Unit 2 Step 4.35.1.e (4.37.1.e in U2 Att. 8 Step Att.7 and 4.38.1.e in U2 Att. 8) also 4.38.1.e includes bullet to close 2-SI-112S, South SI Pump Discharge Drain &

Leakby ShutoffValve Att. 5 Step Establishing LTOP controls for the Aux. 587' No - There are other means 4.35.1.£3, Safety Injection System. Option to of establishing the SI System Att. 5 Step Open SI Pump Discharge Header L TOP conditions.

4.35.1.£5 Valve power supplies (U2, Step

• 1-ABV-D-R5C/2-ABV-D-R1C for 4.35.1.g) ICM-260 Att. 6 Step

• ABV-A-Rl C for ICM-265 4.35.1.f.3 Additionally Close the Discharge Att. 6 Step Header Valve Between the Seat 4.35.1.f.5 (U2, Equalization Line Shutoff Valve 3/4 Step 4.35.1.g)

• 1-SI-119N Att. 6 Step

• 1-SI-119S 4.35.2.d.3 ******************************

Att. 6 Step Unit 2 Step 4.35.1.g (Att. 7, step 4.35.2.d.5 4.37.1.g and Att. 8 step 4.38.1.g)

Att. 7 Step closes SI Pump Discharge Header 4.35.1.f.3 (U2, Shutoff Valves instead of the Step 4.37.1.f.4) Between the Seat Equalization Line Shutoff Valves .

Att. 7 Step 4.35.1.f.5 Page 231 of 236 INFORMATION USE

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases (U2, Step

• 2-SI-205, North SI Pump 4.37.1.g) Discharge Header Shutoff Valve Att. 8 Step

• 2-SI-206, South SI Pump 4.38.1.f.3 Discharge Header Shutoff Valve Att. 8 Step 4.38.1.f.5

.Att. 8 Step 4.38.1.d.3 Att. 8 Step 4.38.1.d.5 Att. 5 Step 4.36.2 In preparation for placing RHR in Aux. 573' No - Isolating RHR from the (U2, Step 4.36.1) service, Close Recirculation Sump to Recirculation Sump is not Att. 6 Step 4.36.2 the RHR Pump Suction Valves 4 required to achieve CSD.

Att. 7 Step 4.38.2

• RH-104E Att. 8 Step 4.39.2

• RH-104W Att. 5 Step 4.37.3 De-energize the RHR to Upper Aux. 633' No - Not required to achieve Att. 6 Step 4.37.3 Containment Spray ShutoffValves CSD.

4 Att. 7 Step 4.39.3

• AM-D-8B for IM0-330 Att. 8 Step 4.40.3

• AM-A-RIB for IM0-331 Att. 5 Step 4.39 Prepare RHR System for service per NIA 1-0HP-4021-017-002, Att. 6 Step 4.39 1(2)-0HP-4021-017-002, Placing in Placing in Service the Service the Residual Heat Removal Residual Heat Removal Att. 7 Step 4.41 System. 4 System which is reviewed Att. 8 Step 4.42 separately in this document.

Att. 5 Step Establishing Charging Pump Aux. 587' No - Step is for pump 4.40.1.c protection for low flow conditions, protection and does not Att. 6 Step Open Charging Pump ELO Valves impact proceeding to CSD.

4.40.1.c power supplies 4

Att. 7 Step

• ABV-D-5C for QM0-225 4.42.1.c

• ABV-A-5C for QM0-226 Att. 8 Step 4.43.1.c Att. 5 Step Establishing Low Temperature Aux. 587' No -T.S. violation but does 4.41.1.b.l.b, Overpressure Protection (LTOP) not prevent the plant from Att. 5 Step condition for the High Head Safety getting to CSD.

4.41.1.b.2.a, Injection (Charging System) for the desired Charging Pump Att. 5 Step 4 4.41.1.c. l.b,

• CS-300E, East CCP Discharge to RCP Seal Water Injection Filter Att. 5 Step Valve 4A 1.1.c.2.a

• CS-301E, East CCP Discharge Att. 6 Step Valve Page 232 of 236 INFORMATION USE

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Or Att. 6 Step

• CS-300W, West CCP Discharge to 4.41.1.b.2.a,

• RCP Seal Water Injection Filter Att. 6 Step . Valve 4.41.1.c.l.b,

• CS-301W, West CCP Discharge Valve Att. 6 Step 4.41.1.c.2.a Att. 7 Step 4.43.1.b.l.b Att. 7 Step 4.43.1.b.2.a Att. 7 Step 4.43.1.c.1.b Att. 7 Step 4.43 .1.c.2.a Att. 8 Step 4.44.1.b.l.b Att. 8 Step 4.44.1.b.2.a Att. 8 Step 4.44.1.c.l.b Att. 8 Step 4.44.1.c.2.a Att. 5 Step 4.43 .2 Place RHR System in service per NIA 1/2-0HP-4021-017-002, Att. 6 Step 4.43.2 1(2)-0HP-4021-017-002, Placing in Placing in Service the Service the Residual Heat Removal 4 Residual Heat Removal Att. 7 Step 4.45.2 System. System which is reviewed Att. 8 Step 4.46.2 separately in this document.

Att. 5 Step 4.52.6, Initiate 1(2)-0HP-4021-054-008, Turb. 609', 633' No - Long path recirculation Att. 6 Step 4.4.4 Att. l, Long Path Recirculation for in this step is to cooldown the 3rd Bullet Cooldown. feedwater flowpath, therefore it does not impact the ability Att. 7 Step 4.55.7 to achieve CSD.

Att. 8 Step 4.6.4 3rd Bullet Step 4.1.2 Verify Recirc Sump Manual Isolation Aux. 573' No - Isolating RHR from the Valves to RHR Suction Closed 4 Recirculation Sump is not required to achieve CSD.

Step 4.1.3 Close Breaker EZC-C-R2D for ICM- Aux. 609' (4KV No - There are other 111, RHR Normal Cooldown Valve Room, flowpaths available for RHR 4

Mezzanine injection if the normal Area) flowpath is not available.

Step 4.2.1.b OpenRH~I21W, WestRHRHeat Aux. 609 (West No -,- Step is only required if Exchanger to Chemical Volume RHRHeat 4 it is desired to minimize Control System (CVCS)

Page 233 of 236 INFORMATION USE ,.

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Demineralizers Shutoff Valve, if Exchanger (Hx) differential pressure across pressurizing from the West RHR Room) the loop 2 suction valves.

Train:

Step 4.2.2.b Open RH-12 lE, East RHR Heat Aux. 609 (East No - Step is only required if Exchanger to CVCS Demineralizers RHRHxRoom) it is desired to minimize 4

Shutoff Valve, if pressurizing from differential pressure across the East RHR Train the loop 2 suction valves.

Step 4.2.4.a IfDemineralizer operation is not Aux. 587' No - Reactor Coolant Filter allowed the Reactor Coolant Filter is may be left in service.

bypassed with RP and Chemistry approvals* 4 Open CS-380, RC Filter Bypass Close CS-377, RC Filter Inlet Close CS-379, RC Filter Outlet Step 4.3.1.b Isolating the East RHR Train if Aux. 609' (East No - Step only prevents desired requires: RHRHxRoom) heating the RHR piping and

• Close RH-128E, East RHR Pump will not prevent achieving Discharge Cross-tie Valve 4 CSD.

• Close RH-121E, East Hx Outlet To the CVCS Demineralizers Step 4.3.2.b Isolating the West RHR Train if Aux. 609 (West No - Step only prevents desired requires: RHRHxRoom) heating the RHR piping and

• CloseRH-128W, WestRHRPump will not prevent achieving Discharge Cross-tie Valve 4 CSD.

• CloseRH-121W, WestHxOutlet To the CVCS Demineralizers Step 4.5.1 Close the following breakers for the Aux. 609' (4KV :Yes?Jf uiiab~(!;fo aHgri1illR

• RHR suction from the Loop 2 Hot Room, " up .via the hopeg Leg Mezzanine 4 *recirculati<;m it wo.uld prevent.

• EZC-B-4C, for IM0-128 Area) ,the"llqs fr<nfa9hi~vini:

• EZC-C-3C, forICM-129 .CSD ..

• Step 4.7 QC perform applicable confirmatory Aux. Various No - per NOTE prior to step UT inspections for the RHR System 4.7, the UT is not required for per 2-EHP-4030-208-006, Shutdown 12 after entering Mode 4 and 4

Monitoring and Trending of Gas per step 4, the UT may be Accumulation in ECCS. deferred if performing a T.S.

required cooldown.

Step 4.9 Open RH-117, RHR East & West Hx Aux. 609' (East No - The RHR Hx bypass bypass Valve. RHRHxRoom) allows for more reliable and 4 predictable temperature control but would not prevent achieving CSD.

Page 234 of 236 INFORMATION USE I

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases

',Elevation/

s~l,~v~~;:t:~~** **.

,:;;~~6fi,'*,£i;!~

Step 4.10.2 If flushing the West RHR Train to a Aux. 609' (East No - Not Flushing the RHR CVCS HUT then verify &WestRHR system would likely cause

• Closed RH-12IE, East RHR Hx HxRooms) rust and other contaminants to Outlet to CVCS Demineralizers be transported into the RCS 4

as well as elevating the

• OpenRH-121W, WestRHR.Hx Outlet to CVCS Demineralizers dissolved oxygen concentration, but would not prevent achieving CSD ..

Step 4.10.3 If flushing the East RHR Train to a Aux. 609' (East No - Not Flushing the RHR CVCS HUT then verify & WestRHR system would likely cause

• ClosedRH-121W, WestRHR.Hx HxRooms) rust and other contaminants to

. Outlet to CVCS Demineralizers 4.

be transported into the RCS as well as elevating the .

• Open RH-12IE, EastRHR.Hx Outlet to CVCS Demineralizers dissolved oxygen concentration, but would not prevent achieving CSD.

Step 4.10.4 If Flushing the second RHR Hx with Aux. 609' (East No - Not Flushing the RHR an RHR pump running, flush both & WestRHR system would likely cause RHR pumps by opening the following HxRooms) rust and other contaminants to valves be transported into the RCS 4

• RH-128E,EastRHR.Pump as well as elevating the Discharge Crosstie dissolved oxygen

• RH-128W, WestRHR.Pump concentration, but would not Discharge Crosstie prevent achieving CSD.

Step 4.12.1 Verify the RHR Discharge Crosstie Aux. 609' (East No - The RHR pump Valve Open & WestRHR discharge crossties allows for

• RH-128E, EastRHR.Pump HxRooms) RHR Hx bypass for more Discharge Crosstie 4 reliable and predictable temperature control but

• RH-128W, WestRHR.Pump Discharge Crosstie would not prevent achieving CSD.

Step 4.12.2.a Align RHR Letdown Manual Valves Aux. 609' (East No - Inability to establish to establish letdown from the West &WestRHR RHR letdown would not RHR.Loop HxRooms) prevent achieving CSD.

• Verify Open RH-121W, West RHR Hx Outlet to CVCS 4 Demineralizers

• Verify Closed RH-12IE, East RHR Hx Outlet to CVCS Demineralizers Step 4.12.2.b Align RHR Letdown Manual Valves Aux. 609' (East No - Inability to establish to establish letdown from the East & WestRHR RHR letdown would not RHR.Loop HxRooms) prevent achieving CSD.

• Verify Open RH-12IE, East RHR Hx Outlet to CVCS 4 Demineralizers

• Verify Closed RH-121W, West RHR Hx Outlet to CVCS Demineralizers Page 235 of 236 INFORMATION USE

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Step 4.16.1 Placing the Reactor Coolant Filter in Aux. 587' No -Leaving the Reactor

. service if it was previously Bypassed Coolant Filter bypassed may Open Filter Inlet and Outlet Valves . cause some chemistry .

concerns but would not

• CS-377, RC Filter Inlet Valve

• 4 prevent achieving CSD.

• CS-379, RC Filter Outlet Valve Close RC-380, RC Filter Bypass Valve Table R-2 & H-2 Results Table R-2 & H-2 Safe Operation & Shutdown Rooms/Areas Room/Area Mode Applicability Auxiliary Building 573' BART Area 3,4, 5 Auxiliary Building 587' Boric Acid Storage Tank Room, Nuclear 3,4, 5 Sampling Room 4KV Room (Mezzanine Area), Boric Acid Batch Tank Area, Chemistry 3,4, 5 Hot Lab, RHR Heat Exchanger Room

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