LR-N22-0039, Emergency Plan Document Revisions Implemented March 24, 2022
| ML22111A064 | |
| Person / Time | |
|---|---|
| Site: | Salem, Hope Creek  |
| Issue date: | 04/21/2022 |
| From: | Barr S Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N22-0039 EP-AA-120-1001 Revision 5 | |
| Download: ML22111A064 (16) | |
Text
LR-N22-0039 April 21, 2022 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 10 CFR 50.54(q)
Salem Nuclear Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354
Subject:
Emergency Plan Document Revisions Implemented March 24, 2022.
Pursuant to 10 CFR 50.54(q) and 10 CFR 50.4(b)(5), PSEG Nuclear LLC (PSEG) is submitting 10 CFR 50.54(q) Summary Analysis Report numbered 2022-03 for the revision to procedures EP-SA-325-141, Salem Emergency Classification Guide Wall Chart, EP-SA-325-117, Salem Section S - S4 - RCS Activity, and EP-SA-325-217, Salem Section S EAL Technical Basis implemented on March 24, 2022 (Attachments 1, 2, 3 and 4).
There are no regulatory commitments contained in this letter.
Should you have any questions, or require further information regarding this submittal, please contact Ms. Megean M. Brown (856) 339-1773.
Respectfully, Stephen T. Barr Manager, Emergency Preparedness SIGN Barr, Stephen T.
Digitally signed by Barr, Stephen T.
Date: 2022.04.18 07:49:43
-04'00'
LRLRN LR-N22-0039 - 10 CFR 50.54(q) Summary Analysis Report 2022-03 - EP-SA-325-141 - Salem Emergency Classification Guide Wall Chart - EP-SA-325-117 - Salem Section S - S4 - RCS Activity - EP-SA-325-217 - Salem Section S EAL Technical Basis cc (w/ Attachments):
USNRC Administrator, Region I USNRC Project Manager USNRC Senior Resident Inspector, Salem USNRC Senior Resident Inspector, Hope Creek (w/o Attachments):
NJDEP Bureau of Nuclear Engineering PSEG Corporate Commitment Tracking Coordinator
LR-N22-0039 ATTACHMENT 1 10 CFR 50.54(q) Summary Analysis Report 2022-03
EP-AA-120-1001 Revision 5 ATTACHMENT 3 10CFR50.54(q)
SUMMARY
ANALYSIS REPORT Page 1
of 3
Revision 0
50.54Q I.D. Number:
2022-03 50.54Q
Title:
Revision to Emergency Action Level (EAL) SU4.2:
EP-SA-325-117, Rev. 1, RCS Activity (Flow Chart)
EP-SA-325-217, Rev. 1, RCS Activity (Technical Basis)
EP-SA-325-141, Rev. 2, EAL Wall Chart - Hot Conditions (Doc #, Rev. #, Name, If applicable)
Description of the change made to the Emergency Plan/Procedures:
EAL SU4.2, as defined in EP-SA-325-117, EP-SA-325-217 and EP-SA-325-141, is being revised IAW Technical Specification Amendment Nos. 337 and 318 (LAR S20-01). The amendment removed Figure 3.4-1 and associated references from the Technical Specifications for both Salem U1 (TS 3.4.8) and Salem U2 (TS 3.4.9), and inserts a limit of less than or equal to the site-specific Dose Equivalent Iodine (DEI) spiking limit of 60 microcuries per gram. A new specific activity for Dose Equivalent Xe-133 (DEX) is also implemented by this amendment. The Technical Specifications have been modified to provide an action for when DEX is not 600 Ci/gram, and to remove the limit associated with gross activity of the reactor coolant Ebar (). The site-specific limit of 600 Ci/gram DEX is established based on the maximum accident analysis RCS activity corresponding to 1 percent fuel clad defects. IAW with the analysis provided in LAR S20-01, if iodine or noble gas spiking were to occur, the normal coolant concentration would be restored within the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> time period provided. Also, there is a low probability of a Steam Line Break (SLB) or Steam Generator Tube Rupture (SGTR) occurring during this time period.
The EAL will be revised to remove the reference to Technical Specification Figure 3.4-1, and replace with the following EAL Threshold values:
SU4.2 Reactor coolant activity > ANY:
60 Ci/gram Dose Equivalent I-131 1.0 Ci/gram Dose Equivalent I-131 for > 48 hrs.
600 Ci/gram Dose Equivalent XE-133 for > 48 hrs.
Description of why the change is editorial (if not editorial, N/A this block):
N/A
EP-AA-120-1001 Revision 5 ATTACHMENT 3 10CFR50.54(q)
SUMMARY
ANALYSIS REPORT Page 2
of 3
Revision 0
50.54Q I.D. Number:
2022-03 50.54Q
Title:
Revision to Emergency Action Level (EAL) SU4.2:
EP-SA-325-117, Rev. 1, RCS Activity (Flow Chart)
EP-SA-325-217, Rev. 1, RCS Activity (Technical Basis)
EP-SA-325-141, Rev. 2, EAL Wall Chart - Hot Conditions (Doc #, Rev. #, Name, If applicable)
Description of the licensing basis affected by the change to the Emergency Plan/Procedure (if not affected, omit this element):
The following Emergency Plan Sections were reviewed:
Emergency Plan Section 5.0 - Emergency Classification System Emergency Plan Section 6.0 - Notification Methods Emergency Plan Section 10.0 - Accident Assessment Emergency Plan Section 16.0 - Radiological Emergency Response Training The emergency plan sections listed above describe methods and processes for accident assessment, classification and notifications, as well as training requirements. These sections do not specify emergency action thresholds or specific activity limits for reactor coolant and therefore are not impacted by the proposed change.
A description of how the change to the Emergency Plan/Procedures still complies with regulation:
The addition of the specific activity limits (e.g. DEX) is consistent with guidance provided in NEI 99-01, Revision 6 that states: Developers may reword the EAL to include the reactor coolant activity parameter(s) specified in Technical Specifications and the associated allowable limit(s) (e.g., values for dose equivalent 1-131 and gross activity, time-dependent or transient values, etc.). If this approach is selected, all RCS activity allowable limits should be included.
For 10 CFR 50.47(b)(4), Emergency Classification System, Reg. Guide 1.219 states that the following examples would generally not require prior NRC approval:
(1) A change to an EAL numeric threshold to reflect an approved change in a technical specification, provided that the basis of the approved EAL is unchanged (e.g., an EAL basis refers to a particular technical specification but not a limiting condition for operation value), and (2) A change to an EAL numeric threshold to reflect a change in a plant design parameter, instrument response characteristics, or design calculation, provided that the meaning or intent of the basis of the approved EAL is unchanged.
The proposed change complies with 10 CFR 50 Appendix E, Regulatory Guide 1.219, Revision 1, 10 CFR 50.47, and with industry guidance in NEI 99-01, Revision 6.
EP-AA-120-1001 Revision 5 ATTACHMENT 3 10CFR50.54(q)
SUMMARY
ANALYSIS REPORT Page 3
of 3
Revision 0
50.54Q I.D. Number:
2022-03 50.54Q
Title:
Revision to Emergency Action Level (EAL) SU4.2:
EP-SA-325-117, Rev. 1, RCS Activity (Flow Chart)
EP-SA-325-217, Rev. 1, RCS Activity (Technical Basis)
EP-SA-325-141, Rev. 2, EAL Wall Chart - Hot Conditions (Doc #, Rev. #, Name, If applicable)
A description of why the proposed change was not a reduction in the effectiveness of the Emergency Plan/Procedure:
The proposed revision aligns with Technical Specification Amendment Nos. 337 and 318 (LAR S20-01) that affects Technical Specifications for both Salem U1 (TS 3.4.8) and Salem U2 (TS 3.4.9). The EAL change will require training be provided to Operations personnel through the Licensed Operator Training program. Training for the Emergency Coordinators in the TSC and EOF and their Direct Reports will be provided through Emergency Preparedness Focus Area Drills (FADs).
There is no reduction in effectiveness to the Emergency Plan resulting from the proposed change to EAL SU4.2, as defined in EP-SA-325-117, EP-SA-325-217, and EP-SA-325-141.
LR-N22-0039 ATTACHMENT 2 EP-SA-325-141 Salem Emergency Classification Guide Wall Chart
None GENERAL EMERGENCY Implement Att. 4 SITE AREA EMERGENCY Implement Att. 3 ALERT Implement Att. 2 UNUSUAL EVENT Implement Att. 1 (Att. 24 for Common Site)
S System Malfunct.
1 Loss of AC Power 6
RPS Failure 7
Loss of Commun.
Inability to shutdown the reactor causing a challenge to RCS water level or RCS heat removal SU1.1 Loss of ALL offsite AC power to 4 KV vital buses for 15 min. (Note 1)
SA1.1 AC power capability to 4 KV vital buses reduced to a single power source for 15 min. (Note 1)
AND ANY additional single power source failure will result in loss of ALL AC power to SAFETY SYSTEMS SS6.1 An automatic or manual trip did NOT shut down the reactor as indicated by reactor power 5%
AND ALL actions to shut down the reactor are NOT successful as indicated by reactor power 5%
AND EITHER:
CFST Core Cooling RED path conditions met CFST Heat Sink RED path exists due to actual loss of secondary heat sink and heat sink is required Automatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are NOT successful in shutting down the reactor SA6.1 An automatic or manual trip did NOT shut down the reactor as indicated by reactor power 5%
AND Manual trip actions taken at the reactor control console (reactor trip switches, trip bkr bezels, supply breakers 1/2E6D and 1/2G6D) are NOT successful in shutting down the reactor as indicated by reactor power 5% (Note 8)
Loss of ALL onsite or offsite communications capabilities SU7.1 Loss of ALL Table S-3 onsite communication methods OR Loss of ALL Table S-3 offsite communication methods OR Loss of ALL Table S-3 NRC communication methods Reactor coolant activity greater than Technical Specification allowable limits SU4.1 RCS leakage for 15 minutes or longer SU5.1 RCS UNIDENTIFIED or PRESSURE BOUNDARY LEAKAGE > 10 gpm for 15 min.
OR RCS IDENTIFIED LEAKAGE > 25 gpm for 15 min.
OR Leakage from the RCS to a location outside containment
> 25 gpm for 15 min.
(Notes 1, 11)
Automatic or manual trip fails to shut down the reactor SU6.1 An automatic or manual trip did NOT shut down the reactor after ANY RPS setpoint is exceeded or a manual trip action was initiated AND A subsequent automatic trip or manual trip action taken at the reactor control console (reactor trip switches, trip bkr bezels, supply breakers 1/2E6D and 1/2G6D) is successful in shutting down the reactor as indicated by reactor power < 5% (Note 8)
Prolonged loss of ALL offsite and ALL onsite AC power to vital buses Loss of ALL offsite power and ALL onsite AC power to vital buses for 15 minutes or longer SS1.1 Loss of ALL offsite and ALL onsite AC power to 4 KV vital buses for 15 min. (Note 1)
SG1.1 Loss of ALL offsite and ALL onsite AC power to 4 KV vital buses AND EITHER of the following:
Restoration of at least one vital bus in < 4 hrs is NOT likely (Note 1)
CFST Core Cooling RED path conditions met Loss of ALL vital DC power for 15 minutes or longer SS2.1
< 114 VDC bus voltage indications on ALL 125 VDC vital buses for 15 min.
< 25 VDC bus voltage indications on both 28 VDC vital buses for 15 min.
(Note 1)
UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress SA3.1 UNPLANNED loss of Control Room indications for 15 minutes or longer SU3.1 Letdown Line Monitor readings indicating fuel clad degradation based on receipt of EITHER of the following (Note 11):
1R31A in warning 2R31 in alarm None EAL WALL CHART - HOT CONDITIONS (RCS > 200°F)
An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for 15 min. (Note 1)
An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for 15 min.
(Note 1)
AND ANY significant transient is in progress, Table S-2 Loss of ALL but one AC power source to vital buses for 15 minutes or longer Loss of ALL offsite AC power capability to vital buses for 15 minutes or longer Reactor coolant activity > ANY: (Note 11) 60 µCi/gram DOSE EQUIVALENT I-131 1.0 µCi/gram DOSE EQUIVALENT I-131 for > 48 hrs.
600 µCi/gram DOSE EQUIVALENT XE-133 for > 48 hrs.
Loss of DC Power 3
Loss of CR Indications None None SU4.2 Loss of ALL vital AC and vital DC power sources for 15 minutes or longer SG2.1 Loss of ALL offsite and ALL onsite AC power to 4 KV vital buses for 15 min.
AND EITHER:
< 114 VDC bus voltage indications on ALL 125 VDC vital buses for 15 min.
< 25 VDC bus voltage indications on both 28 VDC vital buses for 15 min.
(Note 1)
RCS Activity 5
RCS Leakage Hazardous event affecting SAFETY SYSTEMS needed for the current operating mode SA9.1 Table S-4 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 Shift Manager 9
Hazardous Event Affecting Safety Systems None None None None None None None None 2
3 4
1 2
3 4
1 2
3 4
1 2
3 4
1 2
3 4
1 2
3 4
1 2
3 4
1 2
3 4
1 2
3 4
1 2
3 4
1 2
3 4
1 2
1 2
1 2
1 2
3 4
1 2
3 4
1 Failure to isolate containment or loss of containment pressure control SU8.1 ANY penetration is not isolated within 15 min. of a VALID containment isolation signal OR Containment pressure > 15 psig with < one full train of containment depressurization equipment operating per design for 15 min.
(Notes 1,10) 2 3
4 1
None None None 8
CMT Failure
Reactor power
RCS level
RCS pressure
CET temperature
Level in at least one SG
Auxiliary or emergency feedwater flow to at least one SG Table S-1 Safety System Parameters
Automatic turbine runback > 25%
thermal reactor power
Electrical load rejection > 25% full electrical load
Safety Injection Activation Table S-2 Significant Transients Use of Fission Product Barrier Table A point system is used to determine the Emergency Classification Level based on the Fission Product Barrier Table. Each Fission Product Barrier Loss and Potential Loss threshold is assigned a point value as noted below.
Perform the following:
1.
Review all columns of the Fission Product Barrier Table and identify which need further review.
2.
For each of the three barriers, determine the EAL with the highest point value. No more than one EAL should be selected for each barrier.
3.
Add the point values for the three barriers.
4.
Classify based on the point value sum as follows:
FPB Category Salem - Fission Product Barrier Table Inadequate Heat Removal FB1.L CFST Core Cooling RED path conditions met FB1.P CFST Core Cooling PURPLE path conditions met FB2.P CFST Heat Sink RED path exists due to actual loss of secondary heat sink and heat sink is required CMT Radiation /
RCS Activity Loss (5 pts)
Potential Loss (4 pts)
Loss (5 pts)
Potential Loss (4 pts)
None RB2.P CFST Thermal Shock RED path conditions met RB3.P CFST Heat Sink RED path exists due to actual loss of secondary heat sink and heat sink is required CB3.P CFST Containment RED path conditions met CB1.P CFST Core Cooling RED path conditions met AND Restoration procedure 1(2)EOP-FRCC-1 NOT effective within 15 min.
Loss (3 pts)
Potential Loss (2 pts)
None None FB2.L Containment radiation monitor 1(2)R44A or 1(2)R44B reading
> 300 R/hr RB2.L ANY of the following containment radiation monitor readings:
1(2)R2 > 1000 mR/hr 1(2)R44A > 10 R/hr 1(2)R44B > 10 R/hr CB2.P Containment radiation monitor 1(2)R44A or 1(2)R44B reading > 2000 R/hr CMT Integrity or Bypass CB2.L Containment isolation is required AND EITHER:
Containment integrity has been lost based on Emergency Coordinator judgment UNISOLABLE pathway from containment to the environment exists CB3.L Indications of RCS leakage outside of containment CB4.P Containment hydrogen concentration
> 4%
CB5.P Containment pressure > 15 psig with
< one full train of containment depressurization equipment operating per design for 15 min.
(Notes 1, 10)
Reactor Coolant System Barrier Fuel Clad Barrier Containment Barrier FB3.L Coolant activity > 300 µCi/gm dose equivalent I-131 RB3.L ANY condition in the opinion of the Emergency Coordinator that indicates loss of the RCS barrier RB4.P ANY condition in the opinion of the Emergency Coordinator that indicates potential loss of the RCS barrier CB4.L ANY condition in the opinion of the Emergency Coordinator that indicates loss of the Containment barrier CB6.P ANY condition in the opinion of the Emergency Coordinator that indicates potential loss of the Containment barrier FB4.L ANY condition in the opinion of the Emergency Coordinator that indicates loss of the Fuel Clad barrier EC Judgment FB3.P ANY condition in the opinion of the Emergency Coordinator that indicates potential loss of the Fuel Clad barrier None None If sum is:
Classify as:
Emergency Action Levels (EALs)
Refer to ECG ATT#
ALERT 4, 5 ANY loss or ANY potential loss of either Fuel Clad or RCS 2
SITE AREA EMERGENCY 6 -
11 Loss or potential loss of ANY two barriers, OR Potential loss of 2 barriers with the loss of the 3rd barrier 3
GENERAL EMERGENCY 12 or 13 Loss of ANY two barriers AND Loss or potential loss of the third barrier 4
2 3
4 1
MODEs None None None RCS or SG Tube Leakage RB1.P RCS leakage > 50 gpm due to EITHER:
UNISOLABLE RCS leakage SG tube leakage RB1.L An automatic or manual ECCS (SI) actuation required by EITHER:
UNISOLABLE RCS leakage SG tube RUPTURE CB1.L A leaking or RUPTURED SG is FAULTED outside of containment None None None None NOTES Note 1:
The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
Note 10:
One full train of depressurization equipment consists of EITHER:
- at least 5 CFCUs running in low speed with NO Containment Spray train in service
- at least 3 CFCUs running in low speed with one Containment Spray train in service 5.
Implement the appropriate ECG Attachment per above table.
6.
Continue to review the Fission Product Barrier Table for changes that could result in emergency escalation or de-escalation.
Table S-3 Communications Methods System Onsite Offsite NRC Direct Inward Dial System (DID)
X X
X Station Page System (Gaitronics)
X Station Radio System X
Nuclear Emergency Telephone System (NETS)
X X
Centrex Phone System (ESSX)
X X
NRC (ENS)
X The occurrence of ANY Table S-4 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 on 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 12, 13)
Modes:
EAL WALL CHART (HOT)
EP-SA-325-141 Revision 02 1
2 3
4 Power Operations Startup Hot Standby Hot Shutdown NOTES Note 1:
The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.
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.
Note 10: One full train of depressurization equipment consists of EITHER:
at least 5 CFCUs running in low speed with NO Containment Spray train in service at least 3 CFCUs running in low speed with one Containment Spray train in service Note 11: Refer to the Fission Product Barrier Table for possible event escalation due to RCS leakage or high RCS activity.
Note 12: 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 13: 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.
SALEM GENERATING STATION
LR-N22-0039 ATTACHMENT 3 EP-SA-325-117, Salem Section S S4 - RCS Activity
Section S - System Malfunction S4 - RCS Activity Initiating Condition MODE Action Required IF SU4.1 Refer to Attachment 1 UNUSUAL EVENT EP-SA-325-117 Rev. 01 Page 1 of 1 E
M E
R G
E N
C Y
A C
T I
O N
L E
V E
L S
EAL #
THEN Reactor coolant activity greater than Technical Specification allowable limits SGS ECG S4 IF SU4.2 1, 2, 3, 4 Refer to Attachment 1 UNUSUAL EVENT THEN Note 11: Refer to the Fission Product Barrier Table for possible event escalation due to RCS leakage or high RCS activity Letdown Line Monitor readings indicating fuel clad degradation based on receipt of EITHER of the following:
1R31A in warning 2R31 in alarm (Note 11)
Reactor coolant activity > ANY:
60 µCi/gram DOSE EQUIVALENT I-131
1.0 µCi/gram DOSE EQUIVALENT I-131 for > 48 hrs.
600 µCi/gram DOSE EQUIVALENT XE-133 for > 48 hrs.
(Note 11)
LR-N22-0039 ATTACHMENT 4 EP-SA-325-217, Salem Section S EAL Technical Basis
SGS ECG - EAL Technical Bases EP-SA-325-217 Salem Page 1 of 2 Rev. 1 EAL#: SU4.1 EAL Category:
S - System Malfunction EAL Subcategory:
4 - RCS Activity Initiating Condition:
Reactor coolant activity greater than Technical Specification allowable limits Mode Applicability:
1 - Power Operations, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown EAL# & Classification Level:
SU4.1 - UNUSUAL EVENT EAL:
Letdown Line Monitor readings indicating fuel clad degradation based on receipt of EITHER of the following (Note 11):
- 1R31A in warning
- 2R31 in alarm Note 11: Refer to the Fission Product Barrier Table for possible event escalation due to RCS leakage or high RCS activity.
Basis:
This EAL 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 category F or category R ICs.
Explanation/Discussion/Definitions:
Letdown Line Monitors serve as a failed fuel detector by monitoring gamma levels in the reactor coolant letdown line. Unit 1 Letdown Line Monitor (1R31A) and Unit 2 Letdown Line Monitor (2R31) measures letdown line activity. The Letdown Line Monitor warning setpoints are administratively set at 50% of the alarm setpoints.
- 1R31A alarm setpoint is based on 1% failed fuel. The warning setpoint represents about 0.5% failed fuel and has been selected because the setpoint would be readily identifiable on Control Room instrumentation.
SGS ECG - EAL Technical Bases EP-SA-325-217 Salem Page 2 of 2 Rev. 1 EAL#: SU4.1
- 2R31 alarm setpoint is based on 0.1% failed fuel. This setpoint is readily identifiable and also representative of typical values of coolant activity at Technical Specification limits.
Read-outs for these monitors can be obtained in the Control Room.
Other radiation monitors that may be used to confirm a VALID Letdown Line Monitor alarm include:
- 1(2)R4 Charging Pump Room
- 1(2)R26 Reactor Coolant Filter
- Containment Area Rad Monitors (1(2)R2, 1(2)7, 1(2)10A, 1(2)10B)
Definitions:
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 indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment EAL Bases Reference(s):
- 2. PSBP 315733 Radiation Monitoring System Manual, Unit 1
- 3. PSBP 315734 Radiation Monitoring System Control Manual, Unit 2
- 4. UFSAR 9.3.5.3 Safety Evaluation (Failed fuel Detection System)
- 5. UFSAR 11.4 Radiological Monitoring
- 6. S1(S2).OP-AB.RC-0002 (Q) High Activity in the Reactor Coolant System
SGS ECG - EAL Technical Bases EP-SA-325-217 Salem Page 1 of 3 Rev. 1 EAL#: SU4.2 EAL Category:
S - System Malfunction EAL Subcategory:
4 - RCS Activity Initiating Condition:
Reactor coolant activity greater than Technical Specification allowable limits Mode Applicability:
1 - Power Operations, 2 - Startup, 3 - Hot Standby, 4 - Hot Shutdown EAL# & Classification Level:
SU4.2 - UNUSUAL EVENT EAL:
Reactor coolant activity > ANY:
- 60 µCi/gram DOSE EQUIVALENT I-131
- 1.0 µCi/gram DOSE EQUIVALENT I-131 for > 48 hrs.
- 600 µCi/gram DOSE EQUIVALENT XE-133 for > 48 hrs.
(Note 11)
Note 11: Refer to the Fission Product Barrier Table for possible event escalation due to RCS leakage or high RCS activity.
Basis:
This EAL 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 category F or Category R ICs.
Explanation/Discussion/Definitions:
An UNUSUAL EVENT is only warranted when actual fuel clad damage is the cause of the elevated coolant sample (as determined by RCS sample analysis confirmation).
Escalation to an ALERT or higher emergency classification occurs if a sample analysis of reactor coolant activity exceeds 300 µCi/gm DEI-131 via fission product barrier monitoring.
EAL Bases Reference(s):
SGS ECG - EAL Technical Bases EP-SA-325-217 Salem Page 2 of 3 Rev. 1 EAL#: SU4.2
- 2. SGS Technical Specification Section 3.4.8 - Unit 1 Specific Activity
- 3. SGS Technical Specification Section 3.4.9 - Unit 2 Specific Activity
- 4. S1(S2).OP-AB.RC-0002(Q) High Activity in Reactor Coolant System
SGS ECG - EAL Technical Bases EP-SA-325-217 Salem Page 3 of 3 Rev. 1 EAL#: SU4.2 This page intentionally blank