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 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, Barr, Stephen Digitally signed by Barr, Stephen T.
T. Date: 2022.04.18 07:49:43
-04'00' Stephen T. Barr Manager, Emergency Preparedness
LR-N22-0039 LRLRN Attachment 1 - 10 CFR 50.54(q) Summary Analysis Report 2022-03 Attachment 2 - EP-SA-325-141 - Salem Emergency Classification Guide Wall Chart Attachment 3 - EP-SA-325-117 - Salem Section S - S4 - RCS Activity Attachment 4 - 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
GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Implement Att. 4 Implement Att. 3 Implement Att. 2 Implement Att. 1 (Att. 24 for Common Site)
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 Loss of ALL but one AC power source to vital buses for 15 minutes or Loss of ALL offsite AC power capability to vital buses for 15 minutes or longer 15 minutes or longer longer 1
SG1.1 1 2 3 4 SS1.1 1 2 3 4 SA1.1 1 2 3 4 SU1.1 1 2 3 4 Loss of ALL offsite and ALL onsite AC power to 4 KV vital buses Loss of ALL offsite and ALL onsite AC power to 4 KV vital buses for AC power capability to 4 KV vital buses reduced to a single power Loss of ALL offsite AC power to 4 KV vital buses for AND 15 min. (Note 1) source for 15 min. (Note 1) 15 min. (Note 1)
Loss of EITHER of the following: AND AC Power Restoration of at least one vital bus in < 4 hrs is NOT likely (Note 1) ANY additional single power source failure will result in loss of ALL AC CFST Core Cooling RED path conditions met power to SAFETY SYSTEMS Loss of ALL vital AC and vital DC power sources for 15 minutes or Loss of ALL vital DC power for 15 minutes or longer longer SG2.1 1 2 3 4 SS2.1 1 2 3 4 2 Loss of ALL offsite and ALL onsite AC power to 4 KV vital buses for 15 min.
AND
< 114 VDC bus voltage indications on ALL 125 VDC vital buses for 15 min.
OR None None Loss of EITHER: < 25 VDC bus voltage indications on both 28 VDC vital buses for DC Power < 114 VDC bus voltage indications on ALL 125 VDC vital buses 15 min.
for 15 min. (Note 1)
< 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 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress Table S-2 Significant Transients 3 None Automatic turbine None runback > 25%
SA3.1 1 2 3 4 An UNPLANNED event results in the inability to monitor one or more SU3.1 1 2 3 4 An UNPLANNED event results in the inability to monitor one or Loss of CR thermal reactor power Table S-1 parameters from within the Control Room for 15 min. more Table S-1 parameters from within the Control Room for Indications (Note 1) 15 min. (Note 1)
Electrical load rejection > 25% full AND electrical load ANY significant transient is in progress, Table S-2 Reactor Trip Reactor coolant activity greater than Technical Specification allowable Safety Injection Activation Table S-1 Safety System Parameters limits SU4.1 1 2 3 4 Reactor power 4
Letdown Line Monitor readings indicating fuel clad degradation based RCS level on receipt of EITHER of the following (Note 11):
None None None 1R31A in warning RCS RCS pressure 2R31 in alarm Activity SU4.2 1 2 3 4 CET temperature Reactor coolant activity > ANY: (Note 11) 60 µCi/gram DOSE EQUIVALENT I-131 Level in at least one SG S
1.0 µCi/gram DOSE EQUIVALENT I-131 for > 48 hrs.
Auxiliary or emergency feedwater flow 600 µCi/gram DOSE EQUIVALENT XE-133 for > 48 hrs.
to at least one SG RCS leakage for 15 minutes or longer System Malfunct. SU5.1 1 2 3 4 5
RCS UNIDENTIFIED or PRESSURE BOUNDARY LEAKAGE > 10 gpm for 15 min.
None None None OR RCS RCS IDENTIFIED LEAKAGE > 25 gpm for 15 min.
Leakage OR Leakage from the RCS to a location outside containment
> 25 gpm for 15 min.
(Notes 1, 11)
Inability to shutdown the reactor causing a challenge to RCS water Automatic or manual trip fails to shut down the reactor and subsequent Automatic or manual trip fails to shut down the reactor level or RCS heat removal manual actions taken at the reactor control consoles are NOT successful in shutting down the reactor SS6.1 1 2 SA6.1 1 2 SU6.1 1 2 6
An automatic or manual trip did NOT shut down the reactor as indicated An automatic or manual trip did NOT shut down the reactor as An automatic or manual trip did NOT shut down the reactor after by reactor power 5% indicated by reactor power 5% ANY RPS setpoint is exceeded or a manual trip action was initiated AND AND AND RPS Failure ALL actions to shut down the reactor are NOT successful as indicated Manual trip actions taken at the reactor control console (reactor trip A subsequent automatic trip or manual trip action taken at the by reactor power 5% switches, trip bkr bezels, supply breakers reactor control console (reactor trip switches, trip bkr bezels, supply None AND 1/2E6D and 1/2G6D) are NOT successful in shutting down the reactor breakers 1/2E6D and 1/2G6D) is successful in shutting down the EITHER: as indicated by reactor power 5% (Note 8) reactor as indicated by reactor power < 5% (Note 8)
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 Loss of ALL onsite or offsite communications capabilities 7
Table S-3 Communications Methods SU7.1 1 2 3 4 None System Offsite Loss of ALL Table S-3 onsite communication methods None Onsite NRC Loss of OR Commun. Loss of ALL Table S-3 offsite communication methods Direct Inward Dial System (DID) X X X OR NOTES Station Page System (Gaitronics) X Loss of ALL Table S-3 NRC communication methods Station Radio System X Failure to isolate containment or loss of containment pressure control Nuclear Emergency Note 1: The Emergency Coordinator should declare the event promptly upon determining Telephone System (NETS) X X SU8.1 1 2 3 4 that time limit has been exceeded, or will likely be exceeded. Centrex Phone System (ESSX) X X 8 Note 8: A manual trip action is ANY operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does NOT include manually NRC (ENS)
None X ANY penetration is not isolated within 15 min. of a VALID containment isolation signal OR CMT driving in control rods or implementation of boron injection strategies. Containment pressure > 15 psig with < one full train of Failure containment depressurization equipment operating per design for Note 10: One full train of depressurization equipment consists of EITHER: 15 min.
at leastNone 5 CFCUs running in low speed with NO Containment Spray train inNone None (Notes 1,10) service at least 3 CFCUs running in low speed with one Containment Spray train in service Hazardous event affecting SAFETY SYSTEMS needed for the current Note 11: Refer to the Fission Product Barrier Table for possible event escalation due to operating mode RCS leakage or high RCS activity.
Table S-4 Hazardous Events Note 12: If the affected SAFETY SYSTEM train was already inoperable or out of service SA9.1 1 2 3 4 9
before the hazardous event occurred, then emergency classification is NOT The occurrence of ANY Table S-4 hazardous event warranted. Seismic event (earthquake)
AND Event damage has caused indications of degraded performance on Note 13: If the hazardous event ONLY resulted in VISIBLE DAMAGE, with NO Internal or external FLOODING event Hazardous one train of a SAFETY SYSTEM needed for the current operating indications None of degraded performance to at least one train of a SAFETY None mode Event SYSTEM, then this emergency classification is NOT warranted. High winds or tornado strike AND Affecting EITHER:
Safety FIRE Event damage has caused indications of degraded performance Systems on the second train of the SAFETY SYSTEM needed for the current operating mode EXPLOSION Event damage has resulted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating Other events with similar hazard characteristics as mode determined by the Shift Manager (Notes 12, 13)
Use of Fission Product Salem - Fission Product Barrier Table Barrier Table FPB Fuel Clad Barrier Reactor Coolant System Barrier Containment Barrier Category MODEs 1 2 3 4 Potential Loss (4 pts) Loss (5 pts) Potential Loss (4 pts) Loss (5 pts) Potential Loss (2 pts) Loss (3 pts)
A point system is used to determine the Emergency RB1.P RB1.L CB1.L Classification Level based on the Fission Product RCS leakage > 50 gpm due to An automatic or manual ECCS (SI) A leaking or RUPTURED SG is FAULTED Barrier Table. Each Fission Product Barrier Loss and EITHER: actuation required by EITHER: outside of containment RCS or SG Potential Loss threshold is assigned a point value as Tube UNISOLABLE RCS leakage UNISOLABLE RCS leakage noted below. Leakage None None SG tube leakage SG tube RUPTURE None Perform the following:
- 1. Review all columns of the RB2.P Fission Product Barrier CFST Thermal Shock RED path Table and identify which need further review. conditions met
- 2. For each of the three FB1.P FB1.L RB3.P CB1.P barriers, determine the CFST Core Cooling PURPLE path CFST Core Cooling RED path CFST Heat Sink RED path exists due CFST Core Cooling RED path conditions EAL with the highest point value. No more than one conditions met conditions met to actual loss of secondary heat sink met EAL should be selected Inadequate and heat sink is required AND for each barrier. Heat Restoration procedure 1(2)EOP-FRCC-1 FB2.P None None Removal NOT effective within 15 min.
- 3. Add the point values for the CFST Heat Sink RED path exists three barriers. due to actual loss of secondary heat sink and heat sink is required
- 4. Classify based on the point value sum as follows:
FB2.L If Refer RB2.L CB2.P Emergency Containment radiation monitor sum Classify as: to ANY of the following containment Containment radiation monitor Action Levels CMT 1(2)R44A or 1(2)R44B reading is: ECG radiation monitor readings: 1(2)R44A or 1(2)R44B reading > 2000 R/hr (EALs) Radiation / > 300 R/hr ATT# RCS Activity 1(2)R2 > 1000 mR/hr None None None 1(2)R44A > 10 R/hr ANY loss or ANY FB3.L 1(2)R44B > 10 R/hr 4, 5 potential loss of either 2 Coolant activity > 300 µCi/gm dose ALERT Fuel Clad or RCS equivalent I-131 CB3.P CB2.L Loss or potential loss of ANY two barriers, CFST Containment RED path conditions Containment isolation is required OR met AND EITHER:
6- SITE AREA 3 EMERGENCY Potential loss of 2 Containment integrity has been lost 11 barriers with the loss of based on Emergency Coordinator CB4.P the 3rd barrier judgment CMT Containment hydrogen concentration UNISOLABLE pathway from Loss of ANY two Integrity or > 4%
None None None None containment to the environment exists barriers Bypass 12 or AND 4 CB5.P 13 GENERAL Loss or potential loss CB3.L EMERGENCY Containment pressure > 15 psig with of the third barrier Indications of RCS leakage outside of
< one full train of containment depressurization equipment operating per containment design for 15 min.
- 5. Implement the appropriate (Notes 1, 10)
ECG Attachment per above table.
- 6. Continue to review the FB3.P FB4.L RB4.P RB3.L CB6.P CB4.L Fission Product Barrier Table ANY condition in the opinion of the ANY condition in the opinion of the ANY condition in the opinion of the ANY condition in the opinion of the ANY condition in the opinion of the ANY condition in the opinion of the for changes that could result EC Judgment Emergency Coordinator that Emergency Coordinator that Emergency Coordinator that indicates Emergency Coordinator that Emergency Coordinator that indicates Emergency Coordinator that indicates loss in emergency escalation or de-escalation. indicates potential loss of the Fuel indicates loss of the Fuel Clad potential loss of the RCS barrier indicates loss of the RCS barrier potential loss of the Containment barrier of the Containment barrier Clad barrier barrier 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 EAL WALL CHART SALEM (HOT) 1 2 3 4 EAL WALL CHART - HOT CONDITIONS Modes: GENERATING EP-SA-325-141 Power Operations Startup Hot Standby Hot Shutdown (RCS > 200°F) STATION Revision 02
LR-N22-0039 ATTACHMENT 3 EP-SA-325-117, Salem Section S S4 - RCS Activity
SGS ECG Section S - System Malfunction EP-SA-325-117 Rev. 01 S4 - RCS Activity Page 1 of 1 Initiating Condition Reactor coolant activity greater than Technical Specification allowable limits MODE 1, 2, 3, 4 EAL # SU4.1 SU4.2 IF IF E Letdown Line Monitor readings indicating fuel clad Reactor coolant activity > ANY:
M degradation based on receipt of EITHER of the E
R following: 60 µCi/gram DOSE EQUIVALENT I-131 G 1.0 µCi/gram DOSE EQUIVALENT I-131 for > 48 hrs.
E 1R31A in warning 600 µCi/gram DOSE EQUIVALENT XE-133 for > 48 hrs.
N 2R31 in alarm C
Y (Note 11) (Note 11)
A C
T I
O N
L E
V E
L S
THEN THEN Action Refer to Attachment 1 Refer to Attachment 1 Required UNUSUAL EVENT UNUSUAL EVENT Note 11: Refer to the Fission Product Barrier Table for possible event escalation due to RCS leakage or high RCS activity S4
LR-N22-0039 ATTACHMENT 4 EP-SA-325-217, Salem Section S EAL Technical Basis
SGS ECG - EAL Technical Bases EP-SA-325-217 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.
Salem Page 1 of 2 Rev. 1 EAL#: SU4.1
SGS ECG - EAL Technical Bases EP-SA-325-217
- 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 Salem Page 2 of 2 Rev. 1 EAL#: SU4.1
SGS ECG - EAL Technical Bases EP-SA-325-217 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
- 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 Salem Page 2 of 3 Rev. 1 EAL#: SU4.2
SGS ECG - EAL Technical Bases EP-SA-325-217 This page intentionally blank Salem Page 3 of 3 Rev. 1 EAL#: SU4.2