ML16055A230

Response to Request for Additional Information and Emergency Action Level Technical Bases Document, Revision 1, Redline Version. Part 3 of 4

ML16055A230
Person / Time
Site:	Catawba
Issue date:	02/19/2016
From:	Henderson K Duke Energy Carolinas
To:	Office of Nuclear Reactor Regulation
Shared Package
ML16055A223	List: CNS-16-010, Response to Request for Additional Information and Emergency Action Level Technical Bases Document, Revision 1, Clean Version. Part 4 of 4 ML16055A226 ML16055A227 ML16055A228 ML16055A230
References
CNS-16-010, TAC MF6166, TAC MF6167
Download: ML16055A230 (52)

Text

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Introduction Table F-i 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. NCS or SG Tube Leakage B. Inadequate Heat removal C. CMT Radiation / NCS Activity

0. CMT Integrity or Bypass E. Emergency Coordinator Judgment Each category occupies a row in Table F-I 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 B would be assigned "FC Loss B.1," the third Containment barrier Potential Loss in Category D would be assigned "CMT P-Loss D.3," etc.

If a cell in Table F-i 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-I 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-I, 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 ifmultiple 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 ,

potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and NCS barriers and a Potential Loss of the Containment barrier can occur. Barrier ,

IRP/0/A/5000/001 Rev. 1 Page 196 of 247I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Losses and Potential Losses are then applied to the algorithms given in EALs FGI .1, FS1 .1, and FA1 .1 to determine the appropriate emergency classification.

In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the NCS 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 Category A, then B....E.

lRPIOIN15000/100 Rev. 1 Page 197 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Table F-I Fission Product Barrier Threshold Matrix Fuel Clad (FC) Barrier Reactor Coolant System (NCS) Barrier Containment (CMT) Barrier category Loss Potential Loss Loss Potential Loss Loss Potential Loss A ,1 An automatic or manual ECCS (SI) actuation required by ETE:1, CSFSTInetgrity-RED Path 1. Alealing or RUPTURED SG is N NCS or None None

• UNISOLABLE NCS conditions met FAULTED outside of containment Nn SG Tube leakage Leakage *SG tobe RUPTURE

1. CPTCi oln-RNi CSFST Core Cooling-REDAPat Bpath conditions met 1. CSFST Heat Sink-RED Path 1.cS orutinme T otngEDPt ndqae 1. CSFST Core Cooling-RED 2. CSFST Heat Sink-RED Path conditions met Indqae Path conditions met conditions met cniin e None AND None AND Neat AND Restoration procedures not Heat sink is required eftective within 15 min. (Note 1)

Removal Heat sinkis required, C 1. EMF53AIB> Table F-2 column CM CLs~1 EMF53A/E *Table F-2 column NnNoe1. EMF53AIB s Table P-2 colum RadiTio 2 Dosesqiaet1-3" F oln None 1."NCS Loss" NoeNn CMT Potential Loss"

/ NCS acvitdy >3001pCi/gm Activity

1. Containment isolation is required AND EITHER: 1. CSFST Contalnmont-RED Path D Containment integrity has conditions met been Iost based on 2. Containment hydrogen concentretlol CMT None None None None EudmergncyCodiaot

-Injudgmen

3. Containment pressure

• 3 patg.

or Bypass *U NISOLABLE pathway from w~4th n one toll train of contalnen Containment to the ennironment cooling operating per dnsign for exists > 15 main.(Note 1)

2. Indicotmons of NCS leakage outside of containment E 1. Any condition in the opinion of the Emergency Coordinator that

1. Any condition in thie opinion of the Emergency Coordinator that 1 n odto nteoiino

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1, Any condition .Ayodtoitepnoof the opinion n the 1. Any condition in the opinion of 1. Anyconditionintheopinionutfthe EC indicotes lass of the fuel clad Indicates potential loss of the fool the Emergency Coordinator that Emergency Coordinator that the Emergency Coordinator that Emergency Coordinator that Judgment bare ldbririndictaes Ions of the NCS bamrer baeindicotes potential lons of the NCS bareindicates loss of the containment cotimnindicotes potentialba elOSS of the SRP/OA/N50001001 Rev. 1 Page 198 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: 1. NCS or SG Tube Leakage Degradation Threat: Loss Threshold:

INone IRPIO/AI5000/O01 Rev. 1 Page 199 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: 1. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

SNone

[RPI/OIN50001001 lRev. 1I Page 200 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inmidequate Heat Removal Degradation Threat: Loss Threshold:

1. CSFST Core Cooling-RED Path conditions met Definition(s):

None Basis:

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 Operator Aid Computer (OAC) (ref. 1).

This reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.

CNS Basis Reference(s):

1. EP/I (2)/5000/F-0 Critical Safety Function Status Trees

2. EP/I(2)IA/5OOO/FR-C.1 Response to Inadequate Core Cooling

3. EP/I(2)/AI5000/FR-C.2 Response to Degraded Core Cooling

4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A IRPIOIAI50001001 Rev. 1 Page 201 of 247I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: B. Inadequate Heat Removal DegraationThCreat PootentiOalGLoss oniiose

1. CSFST Core Cooling-ORANGE Path conditions met Definition(s):

None Basis:

Critical Safety Function Status Tree (CSFST) Core Cooling-ORANGE path indicates indicates subcooling has been lost and that some fuel clad damage may potentially occur. The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).

This reading indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.

CNS Basis Reference(s):

1. EP/II(2)/5000/F-0 Critical Safety Function Status Trees

2. EP/1(2)/A/5OOOIFR-C.1 Response to Inadequate Core Cooling

3. EP/I1(2)/A/5000/FR-C.2 Response to Degraded Core Cooling

4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.A IRP/O/AI5000/O01 Rev. 1I Page 202 of 247

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 conditions met AND Heat sink is required Definition(s):

None Basis:

In combination with NCS Potential Loss B.1, 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).

The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).

The phrase "and heat sink required" precludes the need for classification for conditions in which NCS 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 2 tells the operator to determine if heat sink is required by checking that NCS pressure is greater than any non-faulted SG pressure and NCS Thot 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 or place ND in service for heat removal. For large LOCA events inside the Containment, the SGs are moot because heat removal through the 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 NCS 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 conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.

CNS Basis Reference(s):

1. EP/I1(2)/5000/ F-0 Critical Safety Function Status Trees

2. EP/I (2)/A/5000/FR-H.1 Response to Loss of Secondary Heat Sink RPO/100/O1Re. age 20 of 247l

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases

3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.B

[RPIOIA15000/O01 Rev. I Page 204 of 247

ATTACHMENT 2 Fission Product Barrier L~oss/Potential Loss Matrix and Bases Barrier: Fuel Clad C. CMT Radiation / NCS Activity Category:

Degradation Threat: Loss Threshold:

1. EMF53A/B > Table F-2 column "FC Loss" Table F-2 Containment Radiation - R/hr (EM F53AIB)

Time After SID FCLs C os CMT Potential (Hrs.) Loss 0-1 550 8.8 5500 1-2 400 '8.4 4000 2-8 160 7.0 1600

>8 100 6.2 1000 Definition(s):

None Basis:

The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the containment high range monitors, EMF53A & B. EMF53A & B are located inside containment. The detector range is approximately 1 to 1E8 R/hr (logarithmic scale). Radiation Monitors EMF53A & B provide a diverse means of measuring the containment for high level gamma radiation. (ref. 1).

The Table F-2 values, column FC Loss represents, based on core damage assessment procedure, the expected containment high range radiation monitor (EMF53A & B) response based on a LOCA, for periods of 1, 2, 8 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown, no sprays and NCS pressure < 1600 psig with -2% fuel failure (ref. 2).

The value is derived as follows:

RPI/OAI5000101 5 Figure 3 Containment Radiation Level vs. T]ime for 100% Clad Damage 1, 2, 8 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown without spray and NCS pressure < 1600 psig x 0.02 (rounded)

(ref. 2).

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300 IJCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since SRP/0/A/5000/001 Rev. I Page 205 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 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 NCS Barrier Loss threshold C.1 since it indicates a loss of both the Fuel Clad Barrier and the NCS Barrier.

Note that a combination of the two monitor readings appropriately escalates the ECL to a Site Area Emergency.

CNS Basis Reference(s):

1. IP/0/3314/004 Radiation Monitoring System RP-2C High Range Process Channel Calibration

2. RP/O/A15000101 5 Core Damage Assessment

3. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.A IRPIoIAI50Iooo1o Rev. I Page 206 of 2471

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad C. CMT Radiation / NCS Activity Category:

Degradation Threat: Loss Threshold:

2. Dose equivalent 1-131 coolant activity > 300 pCi/gm Definition(s):

None Basis:

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. The threshold dose equivalent 1-131 concentration is well above that expected for iodine spikes and corresponds to about 2% fuel clad damage. When reactor coolant activity reaches this level the Fuel Clad barrier is considered lost. (ref. 1).

This threshold indicates that NCS radioactivity concentration is greater than 300 IJCi/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 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

There is no Potential Loss threshold associated with NCS Activity / Containment Radiation.

CNS Basis Reference(s):

1. RP/0/A/5000/01 5 Core Damage Assessment

2. NEI 99-01 CMT Radiation / RCS Activity Fuel Clad Loss 3.B SRP/0/A/5000/001 Rev. 1 Page 207 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: C. CMT Radiation / NCS Activity Degradation Threat: Potential Loss Threshold:

SNone IRP/0/A/5000/001 Rev. 1 Page 208 of 247

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ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases

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SRP/oA150/ooo1o Rev. I Page 209of 2471

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

SNone SRP/0/A/5000/001 Rev. 1 Page 210 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases i Barrier: Fuel Clad Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Fuel Clad barrier Definition(s):

None Basis:

The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier deqradation exists if the degradation will likely occur within two hours 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 monitorinqj 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 EOPs: The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is lost.

CNS Basis Reference(s):

NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A RP/0/A/5000/001 Rev. 1 Page 211 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Fuel Clad Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Fuel Clad barrier Basis:

The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours 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 EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any Other factors that are to be used by the Emergency Coordinator in determining whether the Fuel Clad barrier is potentially lost. The Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

CNS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A IRP/0/A/5000/0O1 Rev. 1I Page 212 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. NCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. An automatic or manual ECCS (SI) actuation required by EITHER:

• UNISOLABLE NCS leakage

• SG tube RUPTURE Definition(s):

UNISOLABLE - An open or breached system line that cannot 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):

• Pressurizer pressure < 1845 psig

• Containment pressure > 1.2 psig This threshold is based on an UNISOLABLE NCS 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 NCS Barrier.

This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE NCS 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 1.A will also be met.

CNS Basis Reference(s):

1. EP/1(2)/AI5OOO/E-O Reactor Trip or Safety Injection

2. EP/1 (2)/AI5000/E-3 Steam Generator Tube Rupture

3. NEI 99-01 ROS or SG Tube Leakage Reactor Coolant System Loss 1l.A IRP/0/A/5000/001 Rev. 1 Page 213 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

1. CSFST Integrity-RED path conditions met Definition(s):

None Basis:

The "Potential Loss" threshold is defined by the CSFST Reactor Coolant Integrity - RED path.

CSFST NCS 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 NCS cooldown under pressure (ref. 1, 2).

This condition indicates an extreme challenge to the integrity of the NCS pressure boundary due to pressurized thermal shock - a transient that causes rapid NCS cooldown while the NCS is in Mode 3 or higher (i.e., hot and pressurized).

CNS Basis Reference(s):

1. EP/1I(2)/A/5000IF-0 Critical Safety Function Status Trees

2. EP/I (2)/A15000/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 .B IRP/O/A15000/O01 Rev. 1 Page 214 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. Inadequate Heat Removal Degradation Threat: Loss Threshold:

NoneI IRPIOIAI5000I100 Rev. 1 IPage 215 of 247

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 conditions met AND 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).

The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (QAC) (ref. 1).

The phrase "and heat sink required" precludes the need for classification for conditions in which NCS 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 2 tells the operator to determine if heat sink is required by checking that NCS pressure is greater than any non-faulted SG pressure and NCS Thot 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 or place ND in service for heat removal. For large LOCA events inside the Containment, the SGs are moot because heat removal through the 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. 1, 2).

This condition indicates an extreme challenge to the ability to remove NCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the NCS 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 because inadequate NCS heat removal may result in fuel heat-up IRP/°/A15°°°/°°O1 Rev. 1 Page 216 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases sufficient to damage the cladding and increase NCS pressure to the point where mass will be lost from the system.

CNS Basis Reference(s):

1. EP/I (2)/5000/F-0 Critical Safety Function Status Trees

2. EP/I(2)/A/5000/FR-H.1 Response to Loss of Secondary Heat Sink

3. NEI 99-01 Inadequate Heat Removal NCS Loss 2.B IRP/o/s0oA1 oo0100 Rev. I Page 217 of 2471

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System C. CMT Radiation!/NCS Activity Category:

D.egr 5AdatioThrat: -2clmn"C Loss"

1. EMF53A/B > Table F-2 column "NCS Loss" Table F-2 Containment Radiation - R/hr (EMF53A/B)

Time After SID CMT Potential (Hs)FC Loss NCS Loss Ls 0-1 550 8.8 5500 1-2 400 8.4 4000 2-8 160 7.0 1600

>8 100 6.2 1000 Definition(s):

N/A Basis:

The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the containment high range monitors, EMF53A & B. EMF53A & B are located inside containment. The detector range is approximately I to 1 E8 R/hr (logarithmic scale). Radiation Monitors EMF53A & B provide a diverse means of measuring the containment for high level gamma radiation. (ref. 1).

The value specified represents, based on core damage assessment procedure RP/0/A/5000/01 5 Figure 1, the expected containment high range radiation monitor (EMF53A &

B) response based on a LOCA, for periods of 1, 2, 8 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown with no fuel failure (ref. 2).

The value is derived as follows:

RP/0/A/5000/015 Figure 1 Containment Radiation Level vs. Time for RCS Release for periods of 1, 2, 8 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown (rounded) (ref. 2).

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 NCS Barrier only.

There is no Potential Loss threshold associated with NCS Activity!/Containment Radiation.

RP/0/A/5000/001 Rev.1IPg 18o 4

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases CNS Basis Reference(s):

1. IP/0/3314/004 Radiation Monitoring System RP-2C High Range Process Channel Calibration

2. RP/0/A/5000/01 5 Core Damage Assessment

3. NEI 99-01 CMT Radiation / RCS Activity NCS Loss 3.A SRP/0/A/5000/001 Rev. 1 Page 219 of 247I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: B. CMT Radiation/ NCS Activity Degradation Threat: Potential Loss Threshold:

None IRP/O/A/5000/O01 Rev. 1 Page 220 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

SNone SRP/0/A/5000/001 Rev. 1 Page 221 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: 0. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

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• RP/0/A/5000/001 Rev. I Page 222 of 2471

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the NCS barrier Definition(s):

None Basis:

The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the NCS barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

aImminent barrier deqradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to the recognition of the inability to reach safety acceptance criteria before completion of all checks.

• Barrier monitorinq 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 EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the NCS Barrier is lost.

CNS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment NCS Loss 6.A RP/0/A/5000/001 Rev. 1 Page 223 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: E. Emergency Coordinator Judgment Degradation Threat: Potential Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the NCS barrier Definition(s):

None Basis:

The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the NCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

a Imminent barrier deqradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to the inability to reach final safety acceptance criteria before completing 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 seguences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the NCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

CNS Basis Reference(s):

1. NEI gg-oi Emergency Director Judgment NCS Potential Loss 6.A SRP/0/A/5000/001 Rev. 1I Page 224 of 247I

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. NCS or SG Tube Leakage Degradation Threat: Loss Threshold:

1. A leaking or RUPTURED SG is FAULTED outside of containment Definition(s):

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.

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 NCS 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 a steam generator is decreasing uncontrollably (part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered 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., NCS activity values) and IC SU5 for the NCS barrier (i.e., NCS 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.

SRP/0/A/5000/001 Rev. I Page 225 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 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 releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.

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 Category R ICs.

The ECLs resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.

Affected SG is FAULTED Outside of Containment?

P-to-S Leak Rate Yes No Less than or equal to 25 gpm No classification No classification Greater than 25 gpm Unusual Event per SU5.1 Unusual Event per SU5.1 Requires operation of charging (makeup) a standby pump (NCS Site Ae mrec AraSmegecype e Alert per FA1.1 BarrierPotentialLoss) Ei.

Requires an automatic or manual SieAaEmrncpr ECCS (SI) actuation (NCS Barrier FSi1.1 Alert per FA1 .1 Loss)

There is no Potential Loss threshold associated with NCS or SG Tube Leakage.

CNS Basis Reference(s):

1. EP/I (2)/A/5000/E-0 Reactor Trip or Safety Injection

2. EPII(2)1A/5000/E-3 Steam Generator Tube Rupture

3. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1I.A IRPIO/A/5000/O01l Rev. 1 Page 226 of 2471

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: A. NCS or SG Tube Leakage Degradation Threat: Potential Loss Threshold:

None IRPIOIAI5000I100 Rev. 1 IPage 227 of 247

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 conditions met AND Restoration procedures not effective within 15 min. (Note 1)

Definition(s):

None Basis:

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 Operator Aid Computer (OAC) (ref. 1).

The function restoration 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 is also evaluated. If core exit thermocouple (TC) readings are greater than 1 ,200°F (ref. 1), Fuel Clad barrier is also lost.

This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the NCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.

The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing and/or if reactor vessel level is increasing. Whether or not the procedure(s) will be effective should be apparent within 15 minutes. The Emergency Coordinator should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.

Severe accident analyses (e.g., NUREG-1 150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.

CNS Basis Reference(s):

1. EP/1 (2)/5000/F-0 Critical Safety Function Status Trees

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases

2. EP/I(2)/AI5000IFR-C.1 Response to Inadequate Core Cooling

3. EP/I(2)IAI5000/FR-C.2 Response to Degraded Core Cooling

4. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.A IRPIO/A150001001 Rev. 1I Page 229 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CMT Radiation/NCS Activity Degradation Threat: Loss Threshold:

SNone SRP/O/N15000/Q100 Rev. 1 Page 230 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: C. CMT Radiation/NCS Activity Degradation Threat: Potential Loss Threshold:

1. EMF53AIB > Table F-2 column "CMT Potential Loss" Table F-2 Containment Radiation - R/hr (EMF53A/B)

Time After SID FCLs C os CMT Potential (Hrs.) Loss 0-1 550 8.8 5500 1-2 400 8.4 4000 2-8 160 7.0 1600

>8 100 6.2 1000 Definition(s):

None Basis:

The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) is monitored by the containment high range monitors, EMF53A & B. EMF53A & B are located inside containment. The detector range is approximately 1 to 1E8 R/hr (logarithmic scale). Radiation Monitors EMF53A & B provide a diverse means of measuring the containment for high level gamma radiation. (ref. 1).

The Table F-2 values, column CMT Potential Loss represents, based on core damage assessment procedure, the expected containment high range radiation monitor (EMF53A & B) response based on a LOCA, for periods of 1, 2, 8 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown, no sprays and NCS pressure < 1600 psig with ~20% fuel failure (ref. 2).

The value is derived as follows:

RPIOIA15000101 5 Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage 1, 2, 8 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after shutdown with no spray and NCS pressure < 1600 psig x 0.20 (rounded)

(ref. 2).

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 NCS Barrier Loss thresholds.

IRP/°/N5°°°/°°l1 Rev. 1I Page 231 of 247

ATTACHMENT Fission Product Barrier Loss/Potential 2Loss Matrix and Bases 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 NCS 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.

CNS Basis Reference(s):

1. IP/0/3314/004 Radiation Monitoring System RP-2C High Range Process Channel Calibration

2. RP/0/A/5000/015 Core Damage Assessment

3. NEI 99-01 CMT Radiation / RCS Activity Containment Potential Loss 3.A RP0A50001Rv.1Pg 22o 4

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

1. Containment isolation is required AND EITHER:

• Containment integrity has been lost based on EC 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:

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 NCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the Emergency 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 NCS 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 NCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases 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 containment 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 existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

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 NCS 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.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold A.I.

CNS Basis Reference(s):

1. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.A Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Loss Threshold:

2. Indications of NCS leakage outside of containment RP/0/A/5000/001 Rev. 1 Page 234 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Definition(s):

None Basis:

ECA-1 .2 LOCA Outside Containment (ref. 1) provides instructions to identify and isolate a LOCA outside of the containment. Potential NCS leak pathways outside containment include (ref. 1,2):

• Residual Heat Removal (ND)

° Safety Injection (NI)

• Chemical & Volume Control (NV)

• RCP seals (NC)

• PZRINCS Loop sample lines (NM)

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 NCS 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 NCS 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 NCS 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.

To ensure proper escalation of the emergency classification, the NCS leakage outside of containment must be related to the mass loss that is causing the NCS Loss and/or Potential Loss threshold A.1 to be met.

CNS Basis Reference(s):

1. EP/I (2)/AI5000/ECA-1 .2 LOCA Outside Containment

2. EP/I(2)/A/5000/E-1 Loss of Reactor or Secondary Coolant

3. NEI 99-01 CMT Integrity or Bypass Containment Loss iRP/01A15000/001 IRev. I Page 235 of 2471

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Figure 1: Containment Integrity or Bypass Examples

• • . " hreshod-

" "Airesolne-.

• , " i release Effluent .0 i_. from Auxiliary Building ._M__nit°rs  :":* Vent RCP Seal Cooling poN50oo0ooI RI Rev. 1 Page 236 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category:

0. CMT Integrity or Bypass DegradaSTiConThet:aPotentiRDalh Losdtos me

1. CSFST Containment-RED path 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 15 psig and represents an extreme challenge to safety function. The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).

15 psig is based on the containment design pressure (ref. 2).

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 NCS 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.

CNS Basis Reference(s):

1. EP/II(2)/AN5000IF-O Critical Safety Function Status Trees

2. UFSAR Section 6.2 Containment Systems

3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.A IRPIOIAI5000I100 Rev. 1I Page 237 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: D. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

2. Containment hydrogen concentration > 6%

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. (ref. 1).

The Containment Hydrogen Purge and Sample System (VY) is used to monitor the hydrogen concentration inside containment after a severe accident involving core damage. Samples of Containment air are obtained via the containment hydrogen/oxygen sample lines to the Post Accident Containment Sample (PACS) panel located in the auxiliary building. Additionally, the containment hydrogen analyzer system continuously monitors the hydrogen concentration inside containment (ref. 1).

The lower limit of deflagration of hydrogen in air is approximately 6% and is the maximum concentration at which hydrogen igniters can be placed in service (ref. 2).

To generate such levels of combustible gas, loss of the Fuel Clad and NCS 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.

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.

CNS Basis Reference(s):

1. UFSAR Section 6.2 Containment Systems

2. EP/1(2)/A/5000/FR-Z.4 Response to High Containment Hydrogen Concentration

3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.B IRP/O/A/5000/O01 Rev. 1 Page 238 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: 0. CMT Integrity or Bypass Degradation Threat: Potential Loss Threshold:

3. Containment pressure > 3 psig with < one full train of containment cooling operating .

per design for > 15 mai. (Notes 1, 10) I Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Note 10: Ifthe loss of containment cooling threshold is exceeded due to loss of both trains of VX-CARF, this EAL only applies if at least one train of VX-CARF is not operating, per design, after the 10 minute actuation delay for greater than or equal to 15 minutes.

Definition(s):

None Basis:

The containment Phase B pressure setpoint (3 psig, ref. 1, 2) is the pressure at which the containment cooling systems should actuate and begin performing their function.

One full train of containment cooling operating per design is considered (ref. 1, 2):

• One train of Containment Air Return Fan System (VX-CARF), and

• One train of Containment Spray System (NS)

Once the Residual Heat Removal system is taking suction from the containment sump, with containment pressure greater than 3 psig and procedural guidance, one train of containment spray is manually aligned to the containment sump. If unable to place one NS train in service or without an operating train of VX-CARF (the CARF with a 10-minute delay) within 15 minutes a potential loss of containment exists. At this point a significant portion of the ice in the ice condenser would have melted and the NS system would be needed for containment pressure control. The potential loss of containment applies after automatic or manual alignment of the containment spray system has been attempted with containment pressure greater than 3 psig and less than one full train of NS is operating for greater than or equal to 15 minutes.

The potential loss of containment also applies if containment pressure is greater than 3 psig and at least one train of VX-CARF is not operating after a 10 minute delay for greater than or equal to 15 minutes. Without a single train of VX-CARF in service following actuation, the potential loss should be credited regardless of whether ECCS is in injection or sump recirculation mode after 15 minutes.

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

ATTACHMENT Fission Product Barrier Loss/Potential 2Loss Matrix and Bases that containment heat removal/depressurization systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.

CNS Basis Reference(s):

1. CNS Technical Specification 3.6.6

2. CNS Technical Specification 3.6.6 Bases

3. CNS Technical Specification 3.3.2

4. UFSAR Section 6.2 Containment Systems

5. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.C IRPIO/AI5000/O01 IRev. 1 Page 240 of 247

ATTACHMENT 2 Fission Product Barrier Loss/Potential Loss Matrix and Bases Barrier: Containment Category: F. Emergency Coordinator Judgment Degradation Threat: Loss Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates loss of the Containment barrier Definition(s):

None Basis:

The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier deqradation exists if the degradation will likely occur within two hours 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 monitoringq 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 EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost.

CNS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A RP/0/AI5000/001 Rev. 1 Page 241 of 247

Fission Product BarrierATTACHMENT Loss/Potential 2Loss Matrix and Bases Barrier: Containment  ;

Category: F. Emergency Coordinator Judgment Degradation Threat: Potential Loss i Threshold:

1. Any condition in the opinion of the Emergency Coordinator that indicates potential loss of the Containment barrier Definition(s):

None Basis:

The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier deqradation exists if the degradation will likely occur within two hours 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 monitoringq 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 EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is lost.

CNS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A IRP/0/A/5000/001 Rev. 1 Page8242 of2471

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases

Background

NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration 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 plantrooms or areas with entry-relatedmode applicabilityidentified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operatingprocedures used for normalplant 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 requiredfor each room or area.

The list should not include rooms or areas for which entry is requiredsolely 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 capabilityto acquire and maintain positive pressure within the Control Room envelope.

The review at CNS was completed using the following Controlling Procedures:

• OP/i1(2)/A/61 00/003 (Controlling Procedure For Unit Operation)

• OP/I1(2)/A/6100/002 (Controlling Procedure For Unit Shutdown)

SRP/0/A/5000/001 Rev. 1 Page 243 of 247

-7;: ,-ii7--:.... °i' '?:-:...............

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases

<, ,, . ,+ , - =, ... .. {f CNS Table R-2 and H-2 Bases A review of station operating procedures identified the following mode dependent in-plant t

actions and associated areas that are required for normal plant operation, cooldown or i shutdown:

- }

i OPI1/AI6100/003 Coordinate with Chemistry Aux*iliary Building (Various Enclosure 4.3 Step and Radwaste while Locations) f 3.19: U2 Encd. 4.3 performing NC System 1No

• Step 3.20 Degas. Radwaste continues Degas OPS thru

• shutdown & cooldown "  : ', ', 7 ,, ' I i

OP/I /A6100/003, Align heater vent orifices Turbine Building (Various Enclosure 4.3, Steps per OP/1 (2)/B162501004 Locations)  !:

3.26, 3.27 and 3.28; (Feedwater Heaters, Vents, N '

U2 Encd. 4.3, Steps Drains and Bleed Systems) 1 No 3.28, 3.29 and 3.30 ,Align VI and SP valves * ,{

associated with CFPTs OP/I11AI61001003, Align Auxiliary Steam to Turbine Building (Various Enclosure 4.2, Step CFPTs. Locations) 1No }

3.11: U2 Endl. 4.2.

Step 3.10 OP/I1&21A61001003, Align "C" Htr Drain Turbine Building (568') i Enclosure 4.2, Steps Pump per 3.13; U2 Endl. 4.2 OP/1(2)IB/62501004 i f Step 3.12 (Feedwater Heaters, INo Vents, Drains and i Bleed Systems) for removal from service ,

OP/i11AI61001003, Plant activities to ensure Turbine Building (594')

Enclosure 4.2, Step Main Turbine Sealing Steam 1No i 3.14; U2 End. 4.2 system responds as Step 3.13 required.

OP/i1&21I61001003, Ensure Moisture Separator Turbine Building (619') ... ....... .' + , ...... .,t, ,,:' >" ;*

Enclosure 4.2, Step Reheater low load valve i J 3.18: U2 Endl. 4.2 operation per INo Step 3.17 O P/I(2)/B/6250/013 (Moisture Separator t Reheater Operation)

OPI11A/61001003 Secure one Main CFPT per Turbine Building (Mainly Enclosure 4.2 Step OP/i (2)1AN6250/001 594') 1N oi i

3.19: U2 Endl. 4.2 (Condensate and *,' * , <o: ,,*,*< ,,< <" '7 * ;,;]

Step 3.18 Feedwater System) .: ,+<; .. .. ," ::,,,<  :+ 7,]

....+ 2*" +7 7 <{',<" 'i, :*{

OPIAI61001003, Secure half Main Outside in Main i Enclosure 4.2, Step Transformer Cooling Fans Transformer yard 1No 3.20; U2 End. 4.2 and oil pumps

-Step 3.19 - _ _ __ _ _I _ _  !°__ " :< + < , o']; , ,," , , "7 "f 7;  :*';:'g*' < '" '<i rP//10001Rv ae24o 4 Rev. 1 Page 244 of 247 I I :*PIOIAI50001001 J

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases OP/1 (2)/A/6100/0 03 Shutdown the Main Turbine Turbine Building and Enclosure 4.2 Step per OP/I1(2)/B/6300/001 Transformer Yard No 3.21: (Turbine generator)

OP/i1(2)1A!61 00/003, Bypass "F" LP heaters Turbine Building (594" LP Enclosure 4.2, Step Htr Panel). 1No i 3.28 &3.29 4 OP/I1(2)/A/6100/003, Transfer of Aux Steam to Service Bldg. (568')

Enclosure 4.2, Step on line Unit per 3.34 OP/O/B/6250/007 A (Auxiliary Steam System INo  !

Align ment) or place Aux Electric Boiler in service .

per OP/1/B/6250/007 B '

(Auxiliary Electric Boilers'li OP/I1(2)/A/61 00/003, Isolate Unit Related Turbine Building. (594') ,

Enclosure 4.2, Step Steam supply to Aux 1No 3.37 Steam Header OP/i1(2)/A/6100/002, Initiate action to reduce Auxiliary Building (577' f Enclosure 4.1, Step VCT pressure per Mechanical Pent. Room) 3.4 OP/I1(2)/6200/001 1, 2, 3 No *i..

(Chemical and Volume Control System)

OP/1(2)/A/6100/002, Align S/G reverse purge. Both Doghouses 1 N Enclosure 4.1, Step N "

3.9 OP/1(2)/A/6100/002, Align CM system flow for Turbine Building (619') 3 N Enclosure 4.1, Step Low Pressure 'cleanupNo  !

3.52 thru Upper Surge Tank. '

OP/I1(2)/A/I6100/002 Shutdown Rod Control Auxiliary Building (594' Enclosure 4.2 or 4.7, System per Electrical Pent Room) 3No Step 3.3 OP/1 (2)/6150/008 (Rod i Control)

OP/I1(2)/A/61 00/002 Chemistry obtains Auxiliary Building (543'f Enclosure 4.2 or 4.7 samples to ensure boron Sample Lab) 3N Step 3.10 concentration good to N '

allow NCS cooldown to begin OP/I1(2)/A/6i00/002 Perform PZR PORVs Auxiliary. Building (577', I Enclosure 4.2 or 4.7, stroke testing per IF performing IWVR Step 3.22 PT/I1(2)/A/4200/023 A Containment 635' as 3No*!

(NC valve Inservice Test) well) .

IRPIOIAI50001001 Rev. 1I Page 245 of 247

ATTACHMENT 3  ; ]

Safe Operation & Shutdown Rooms/Areas Tabies R-2 & H-2 Bases OP/1 (2)/A/6100/002 Support placing N2 Auxiliary Building Enclosure 4.2 or 4.7 Cover gas on NCDT per (Various Locations on Step 3.23 OP/Il(2)/A!6500/01 4 577' & 560')

3 No (Operations Controlled Liquid Waste Systems)

OP/I1(2)IAN6100/002, Removing CLAs from Auxiliary Building (577' & }

}

Enclosure 4.2 or 4.7 service per 560' Ess. MCC Bkrs) 3 N Step 3.31 OP/I (2)/A/6200/009 N (Cold Leg Accumulator Operation).

OP/i1&2/A/6100/002 Remove CAPT and one Auxiliary Building. (577' &

Enclosure 4.2 or 4.7, Motor Driven CA Pump from 560 Ess MCC Bkrs)

Step 3.45.1 service per 4No OP/I1(2)/A/6250/002 (Auxiliary Feedwater SystemV) =

OP/I1(2)/A/6100/O02 Open NCS Loop Suction Auxiliary. Building. (577'

• Enclosure 4.2 or 4.7, Vlvs for train of ND to be & 560' Ess. MCC Bkrs)

Step 3.46.3 placed in service per 4Yes OP/i1(2)/A/6200/004 (Residual Heat removal

__________System).*

OP/i1(2)/A/61001002 Rack out appropriate NI and Auxiliary Building (577' &  !

Enclosure 4.2 or 4.7, NV Pump Motor Bkrs per 560 Electrical Pent Step 3.48.2 OPIOIA/63501010 (Operation Rooms) 4Yes of Station Breakers and i Disconnects)

OPI1(2)IA/61001002 Support placing first train of Auxiliary Building (577' or~i /

Enclosure 4.2 or 4.7 ND in service per 560' Ess MCC Bkr s) Ye Step 3.52.2 OP/I (2)1N/620010044Ye, i (Residual Heat removal System) ___________________________

, ' )

, }

IRPIoIA/5000/001 IRev. I Page 246 of 2471

ATTACHMENT 3 Safe Operation & Shutdown Rooms/Areas Tables R-2 & H-2 Bases Table R-2 & H-2 Results Table R-2/H-2 Safe Operation & Shutdown Rooms/Areas Bldg. Elevation Unit I RoomlArea Unit 2 RoomlArea Mode Rrm 478 (1EMXA) Rm 469 (2EMXA) 4 Rm 496 (1ETA) Rm 486 (2ETA) 4 Auxiliary 577' Rm 496 (1EMXS) Rm 486 (2EMXS) 4

________AB-577', J J-57 (1MXK) AB-577', J J-57 (2MXK) 4 Rm 330 (1EMXJ) Rm 320 (2EMXJ) 4 Auxiliary 560' Rm 372 (1 ETB) Rm 362 (2ETB) 4 Rm 372 (1EMXD) Rm 362 (2EMXD) 4 Plant Operating Procedures Reviewed

1. OP/i1(2)/A/6100/003 (Controlling Procedure for Unit Operation)

2. OP/I1(2)/A/6100/002 (Controlling Procedure for Unit Shutdown)

3. OP/I (2)/B/6250/004 (Feedwater Heaters, Vents, Drains and Bleed Systems)

4. OP/i1(2)/B/6250/013 (Moisture Separator Reheater Operation)

5. OP/i1(2)/A/6250/001 (Condensate and Feedwater System)

6. OP/i1(2)/B/6300/O001 (Turbine generator)

7. OP/0/B/6250/007 A (Auxiliary Steam System Alignment)

8. OP/I1/B/6250/007 B (Auxiliary Electric Boilers)

9. OP/i1(2)/6200/001 (Chemical and Volume Control System)

10. OP/I (2)/6150/008 (Rod Control)

11. PT/i (2)/A/4200/023 A (NC valve Inservice Test)

12. OP/i1(2)/A/6500/014 (Operations Controlled Liquid Waste Systems)

13. OP/i1(2)/A/6200/009 (Cold Leg Accumulator Operation)

14. OP/i (2)/A/6250/002 (Auxiliary Feedwater System)

15. OP/i1(2)/A/6200/004 (Residual Heat removal System)

16. OP/O/A/6350/01 0 (Operation of Station Breakers and Disconnects)

IRP/O/A/5000/O01 Rev. 1 Page 247 of 247