NLS2025035, Application to Revise Technical Specifications to Adopt TSTF-576, Revise Safety/Relief Valve Requirements
| ML25237A251 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 08/23/2025 |
| From: | Dia K Nebraska Public Power District (NPPD) |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| NLS2025035 | |
| Download: ML25237A251 (1) | |
Text
N Nebraska Public Power District "Alwqs therc u'lvn Jou nectl us" 50.90 N15202503s August 23,2025 Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Subject:
Application to Revise Technical Specifications to Adopt TSTF-576, "Revise Safety/Relief Valve Requirements" Cooper Nuclear Station, Docket No. 50-298, Renewed License No. DpR-46 Pursuant to 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Nebraska Public Power District (NPPD) is submitting a request for an amendment to the Technical specifications (TS) for cooper Nuclear station (cNS)
NPPD requests adoption of Technical Specification Task Force (TSTF) traveler TSTF-526, "Revise Safety/Relief Valve Requirements." The proposed change revises the Safety/Relief V-alve (SRV) and Safety Valve (SV) TS to align the overpressure protection requirements with the safety limits and the regulations. provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked to show the proposed changes. Attichment 3 provides revised (clean) TS pages. Attachment 4 provides the existing TS Bises pages marked to show revised text associated with the proposed TS changes and is provided tor intormation only.
Attachmeht 5 provides an example of the Core Operating Limits Report, for information, that illustrates the addition of the SRV and SV limits.
NPPD requests that the amendment be reviewed under the Consolidated Line ltem lmprovement Process. Approval of the proposed amendment is requested within six months of acceptance. Once approved, the amendment shall be implemented within 60 days.
The proposed TS change has been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review ani nuoii Board). ln accordance with 10 CFR 50.91, "Notice for public comment; State consultation," a copy of this application, with attachments, is being provided to the designated State of Nebraska Cjfficiat.
NPPD has determined there are no significant hazards considerations associated with the proposed TS change and it qualifies for categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).
COOPER NUCLEAR STATION 72676 648A Ave / P.O. Box g8 / Brownviile, NE 68321 http://v*rwv.nppd.com
There are no regulatory commitments made in this submittal. lf you should have any questions regarding this submittal, please contact Linda Dewhirst, Relutatory Affairs and Compliance Manager, at (aO\\ 82S-S41 6.
I declare under penarty of perjury that the foregoing is true and correct N152025035 Page 2 of 2 Executed On:
Since lDia Site Vice President
/bs Attachments: 1.
2.
3.
4.
cc Date 5
DESCRIPTION AND ASSESSMENT PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)
REVISED TECHNICAL SPECI FICATION PAGES PROPOSED TECHN ICAL SPECI FICATION BASES CHANG ES (MARK-UP)
- FOR INFORMATION ONLY EXAMPLE UPDATED CORE OPERATING LIMITS REPORT - FOR INFORMATION ONLY Regional Ad m inistrator w/ attachments USNRC - Region tV Cooper Prolect Manager w/ attachments USNRC - NRR Plant Licensing Branch tV Senior Resident lnspector w/ attachments USNRC - CNS Nebraska Health and Human Services M attachments Department of Regulation and Licensure NPG Distribution M attachments CNS Records M attachments
N152025035 Page 1 ofS 1.0 2.0
3.0 DESCRIPTION
AND ASSESSMENT Cooper Nuclear Station, Docket No. 50-298, Renewed Operating License No. DpR-46 DESCRIPTION ASSESSMENT 2.1 Applicability of Safety Evaluation 2.2 Optional Changes and Variations REGULATORY ANALYSIS 3.1 No Significant Hazards Consideration Analysis 3.2 Conclusion ENVIRONMENTAL EVALUATION 4.0
N152025035 Page 2 of 5
1.0 DESCRIPTION
Nebraska Public Power District (NPPD) requests adoption of Technical Specification Task Force (TSTF) traveler TSTF-576, "Revise Safety/Relief Valve Requirements." The proposed change revises the Safety/Relief Valve (SRV) and Safety Valve (SV) Technical Specifications (TS) to align the overpressure protection requirements with the safety limits and the regulations
2.0 ASSESSMENT
2.1 Applicability of Safety Evatuation NPPD has reviewed the safety evaluation for TSTF-576 provided to the Technical Specifications Task Force in a letter dated September 10,2024. This review included a review of the Nuclear Regulatory Commission (NRC) staff's evaluation, as well as the information provided in TSTF-576. As described herein, NPPD has concluded that the justifications presented in TSTF-S76 and the safety evaluation prepared by the NRC staff are applicable to Cooper Nuclear Station (CNS) and justify this amendment for the incorporation of the changes to the CNS TS.
The NRC-approved overpressure protection analysis methodology for CNS is NEDE-24011-p-A, "General Electric Standard Application for Reactor Fuel."
2.2 Optional Changes and Variations NPPD is proposing the following variations from the TS changes described in TSTF-576, Revision 3, or the applicable parts of the NRC staff's safety evaluation:
TSTF-576 only refers to SRVs. The same overpressure protection function is performed. Therefore, the justification of TSTF-576 is applicable to CNS.
- 2. The CNS TS utilize different numbering than the Standard Technical Specifications (STS) on which TSTF-576 was based. Specifically, TSTF-576 revises STS LCO Bases 3.4.11, "Reactor Steam Dome Pressure" where that section in the CNS TS Bases is LCO 3.4.10. Also, TSTF-576 revises the STS Bases for Surveillance Requirement (SR) 3.3.6.3.7 where that SR in the CNS TS Bases is SR 3.3.6.3.5.
This difference is administrative and does not affect the applicability of TSTF-576 to the CNS TS.
The CNS TS contain requirements for SRVs and SVs that differ from the STS on which TSTF-576 is based; however, the changes to the CNS TS are encompassed in the TSTF-576 justification. Condition C that is revised in STS LCO 3.4.3 is analogous to condition A in the cNS TS. The cNS TS does not contain any Conditions corresponding to Condition A or B of STS LCO 3.4.3. This difference is administrative and does not affect the applicability of TSTF-576 to the CNS TS.
- 4. The CNS TS SR 3.4.3.1does not contain the Note that is being deleted by TSTF-576. The remaining changes to sR 3.4.3.i are appliable to the cNS TS. This difference is administrative and does not affect the applicability of TSTF-S76 to the CNS TS.
NLS2025035 Page 3 of 5
- 5. The revised STS LCO 3.4.3 Bases Background section in TSTF-576 describes the relief mode operation of an SRV. This discusses the operation of a pneumatic piston and mechanical linkage assembly that is used to open the SRV, even with the valve inlet pressure at 0 psig. The CNS SRV design is an air diaphragm and requires 50 psig to overcome spring force. Accordingly, NPPD will use the CNS-specific SRV design and the 50 psig in the CNS TS LCO 3.4.3 Bases Background section to describe the operation of the SRV. This difference does not affect the applicability of TSTF-576 to the CNS TS.
NPPD reviewed the NRC's request for additional information (ML25042A041) associated with Browns Ferry Nuclear (BFN) license amendment request (M124344A034) to adopt TSTF-576. This review identified that RAI-4 was appticabte to CNS. The proposed CNS TS SR 3.4.3.1list eight SRVs and three SVs white the proposed CNS TS LCO 3.4.3 Bases changes state the overpressure protection analyses assume seven sRVs and three SVs operate in the safety mode of operation. Accordingly, a variation to TSTF-576, like the variation in BFN's response (M125069A477),is proposed. The following additionalwording is being added to the cNS TS Bases for sR 3.4.3.2 to ensure that only the sRVs and sVs that are determined to be operable by as-left testing per SR 3.4.3.1 are credited when performing SR 3.4.3.2:
"Any valve not credited for SR 3.4.3.1may not be credited for this Surveillance in the same operating cycle."
- 7. TSTF-576 revises the STS Bases for sR 3.3.6.3.7 (equivalent to cNS is sR 3.3.6.3.5) by renaming the title of LCo 3.4.3. For the cNS TS Bases for sR 3.3.6.3.5, the reference to LCO 3.4.3 is deleted because with the removal of the manual opening requirement sR from cNS TS Lco 3.4.3, cNS TS LCo 3.4.3 will not perform a part of the logic system functional test of SR 3.3.6.3.5. The manual opening is still performed by CNS SR 3.6.1.6.1, therefore this difference is administrative and does not affect the applicability of rsrF-sz6 to cNS.
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Analysis Nebraska Public Power District (NPPD) requests adoption of Technical Specification fask Force (TSTF) traveler TSTF-576, "Revise Safety/Relief Valve Requirements." The proposed change revises the Safety/Relief Valve (SRV) and Safety Valve (SV) Technical Specifications (TS) to align the overpressure protection requirements with the safety limits and the regulations.
The Limiting Condition for Operation (LCO) is revised to replace requirements on each credited SRV and SV with a requirement that the Overpressure Protection System (OPS) be operable.
The Surveillance Requirements (SRs) are revised to move the as-found SRV and SV lift pressure limits to the licensee-controlled Core Operating Limits Report (COLR). An SR that tests the ability of the SRVs to be capable of manual operation is removed as that capability is not credited in any safety analysis. The TS Actions are revised to be consistent with the changes to the LCO and SRs. Administrative changes are made to the TS for clarity and consistency. The COLR specification is revised to reference the OPS specification.
6
1 N1S2025035 Page 4 of 5 NPPD has evaluated if a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "lssuance of amendment," as discussed below:
Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed change revises the SRV and SV TS to align the overpressure protection requirements with the safety limits and the regulations. The OPS must accommodate the most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSlVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position). The proposed change does not affect the MSIVs and would have no effect on the probability of the MSIVs closing or generation of a reactor scram signal on MSIV position. Therefore, the probability of the event is unaffected. The consequences of the accident are based on the peak reactor pressure vessel pressure. Both the current and proposed TS ensure the overpressure Safety Limit is not exceeded. The accident analyses consider the aggregate operation of the credited SRVs and SVs, not the performance of individual valves. The proposed change moves the SRV and SV as-found lift pressure limits to the licensee-controlled COLR which uses Nuclear Regulatory Commission (NRC) approved methodologies. Altering the control process for these values has no effect on the accident evaluations. As a result, the consequences of the accident are not changed.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Does the proposed amendment create the possibility of a new or different kind of accident,from any accident previously evaluated?
Response: No The proposed change revises the SRV and SV TS to align the overpressure protection requirements with the safety limits and the regulations. The proposed change does not alter the design function or operation of the SRVs or SVs. The proposed change does not create any new credible failure mechanisms, malfunctions, or accident initiitors not already considered in the design and licensing basis.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No The proposed change revises the SRV and SV TS to align the overpressure protection requirements with the safety limits and the regulations. The proposed change ensures that the SRVs and SVs can protect Safety Limit 2.1.2. Although the as-found SRV and SV lift pressure limits are moved to the licensee-controlled COLR, the safety margin 2
3
N1S2025035 Page 5 of 5 provided by the SRVs and SVs, which ensures the Safety Limit is protected, is not changed. The conservatisms in the evaluation and the analysis are described in the NRC-approved methods for each licensee, which are not altered by the proposed change. The proposed change does not alter a design basis limit or a safety limit, and, therefore, does not reduce the margin of safety.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, NPPD concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, aciordingly, a finding of "no significant hazards consideration" is justified.
3.2 Conclusion ln conclusion, based on the considerations discussed above, (1)there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
4,0 ENVIRONMENTAL EVALUATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that.may be released offsite, or (iii) a significant increase in individual or cumulative occupationai radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 cFR 51.22(c)(g). Therefore, pursuant to 10 cFR 51.22(b), ni environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
N152025035 Page 1 ofS PROPOSED TECHNICAL SPECIFICATIONS CHANGES (MARK-UP)
Cooper Nuclear Station, Docket No. 50-298, Renewed Operating License No. DpR-46 Revised paoes il 3.4-6 3.4-7 5.0-21
TABLE OF CONTENTS (continued) 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 3.4.7 3.4.8 3.4.9 3.4.10 3.5 3.5.1 3.5.2 3.5.3 REACTOR COOLANT SYSTEM (RCS).
Recirculation Loops Operating Jet Pumps.
_ gotection Svstem fOru RCS Operational LEAKAGE RCS Leakage Detection lnstrumentation RCS Specific Activity Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown...........
Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown.
RCS Pressure and Temperature (p/T) Limits........
Reactor Steam Dome Pressure EMERGENCY CORE COOLTNG SYSTEMS (ECCS), RpV WATER INVENTORY CONTROL, AND REACTOR CbNC ISOMTIOru cooltNG (RCIC) SYSTEM ECCS - Operating RPV Water lnventory Control RCIC System CONTAINMENT SYSTEMS.........
Primary Containment..............
Primary Containment Air Lock..........
Primary Containment lsolation Valves (pClvs)
Drywell Pressure...
DrywellAir Temperature Low-Low Set (LLS) Vatves......
Reactor Building-to-suppression chamber Vacuum Breakers Suppressio n-Cha m berto-Drywel I Vacu u m Brea ke rs Residual Heat Removal (RHR) Containment Spray Suppression Pool Average Temperature.....................
Suppression Pool Water Level.........
Residual Heat Removal (RHR) Suppression pool Cooling......
Primary Containment Oxygen Concentration............
Secondary Containment..............
Secondary Containment lsolation Valves (SClvs).....
Standby Gas Treatment (SGT) System......
PLANT SYSTEMS Residual Heat Removal service water Booster (RHRSWB) system.
service water (SW) system and Urtimate Heat sink (uHSi....:..........
Reactor Equipment Cooting (REC) System.......................
Control Room Emergency Filter (CREF) System Air Ejector Offgas.......
Spent Fuel Storage Pool Water Level........
The Main Turbine Bypass System 3.4-1 3.4-1 3.4-4
.3.4-6
.3.4-8
.3.4-10
.3.4-13 3.4-15 3.4-18 3.4-20 3.4-24 3.5-1 3.5-1 3.5-7 3.5-12 3.6 3.6.1.1 3.6.1.2 3.6.1.3 3.6.1.4 3.6.1.5 3.6.1.6 3.6.1.7 3.6.1.8 3.6.1.9 3.6.2.1 3.6.2.2 3.6.2.3 3.6.3.1 3.6.4.1 3.6.4.2 3.6.4.3
.....3.6-1
.....3.6-1
,....3.6-3
....3.6-8
....3.6-16
....3.6-17
....3.6-18
....3.6-20
....3.6-23
....3.6-25
....3.6-27
....3.6-30
....3.6-31
....3.6-33
....3.6-34
....3.6-36
....3.6-40 3.7 3.7.1 3.7.2 3.7.3 3.7.4 3.7.5 3.7.6 3.7.7 (continued Cooper Amendment No. 263 I
SR\\ls-e{VsOPS I 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 (oPS)
LCO 3.4.3 APPLICABILITY: MODES 1,2, and 3 ACTIONS CONDITION A. One er mere required SRVs FVsOPS inoperable.
re Protection m
The SVsOpS shail be OpEMBLE.
COMPLETION TIME 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hours REQUIRED ACTION A.1 AND 4.2 Be in MODE 3 Be in MODE 4.
Cooper 3.4-6 Amendment No. 240 I
SURVEILLANCE REQU IREMENTS sR 3.4.3.1 Number of OPS SRVs SURVEILLANCE VerifYthery"*
se+peints,pressu reg of the req u i red SafetL Re lief Valves and Valves ers
! 1o/o ot SRV-ndVsOPS 3.4.3 FREQUENCY ln accordance with the INSERVICE TESTING PROGRAM ln accordance with theuryeifianee Srcguenye+trl Pregram INSERVICE TESTING PROGRAM Nominal Setpoint (psig) 1080+32,4 1090-+33J 1 100-+33 Number of OPS SVs Nominal Setpoint
)
3 1240 +7+
Fellewing testing, lift settings shall be within + 1%,
sR 3.4.3.2 are adequate te pe*erm-t+e+es+
Verify eaeh SRV epens when manually aetuated, Verifv the as-found OPS lift oressures of the reouired SRVs and are within the limits s in the COLR.
2 3
3 Cooper 3.4-7 Amendment No.29 I
5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives ouilined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix l, Section lV.B.1.
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.4 (Deleted) 5.6.5 Core Operatinq Limits Report (COLR)
- a.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
- 1.
The Average Planar Linear Heat Generation Rates for Specifications 3.2.1 and3.7.7.
2-The Minimum Critical Power Ratio for Specificatio ns 3.2.2and 3.7.7, and M CPRgg gv" for Specific ation 3.2.2.
- 3.
The Linear Heat Generation Rates for Specifications 3.2.3 and 3.7.1.
- 4.
The three Rod Block Monitor Upscale Allowable Values for Specifica1on 3.3.2.1.
- 5.
The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.
- 6.
The Minimum Critical Power Ratios in Table 3.3.2.1-1for Specification 3.3.2.1.
- 7.
The as-found ure on Soecification 4.3 b
for The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1 NEDE-24011-P-A, "General Electric standard Apprication for Reactor Fuel" (Revision specified in the COLR).
(continued )
Cooper 5.0-21 Amendment No.278 I
N1S2025035 Page 1 of 5 REVISED TECHNICAL SPECIFICATIONS PAGES Cooper Nuclear Station, Docket No. 50-298, Renewed Operating License No. DpR-46 Revised Paoes il 3.4-6 3.4-7 5.0-21
TABLE OF CONTENTS (continued) 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 3.4.7 REACTOR COOLANT SYSTEM (RCS).......
Recirculation Loops Operating Jet Pumps Overpressure Protection System (OpS)...............
RCS Operational LEAP(AGE............
RCS Leakage Detection lnstrumentation RCS Specific Activity Residual Heat Removal(RHR) Shutdown Cooling System - Hot Shutdown...........
Residual Heat Removal(RHR) Shutdown Cooling System - Cold Shutdown..
RCS Pressure and Temperature (p/T) Limits........
Reactor Steam Dome Pressure EMERGENCY CORE COOLTNG SYSTEMS (ECCS), Rpv WATER INVENTORY CONTROL, AND REACTOR CbNr IsOMTIOT cooLtNG (RClC) SYSTEM ECCS - Operating RPV Water lnventory Control RCIC System CONTAINMENT SYSTEMS.........
Primary Containment..............
Primary Containment Air Lock.
Primary Containment lsolation Valves (pClvs)
Drywell Pressure...
Drywell Air Temperature Low-Low Set (LLS) Vatves Reactor Buildingto-suppression cham ber Vacuum Breakers S u ppress ion-Cha m ber-to-Drywe I I Vacu u m B rea kers Residual Heat Removal(RHR) Containment Spray Suppression Pool Average Temperature.........'...............
Suppression Pool Water Level.........
Residual Heat Removal (RHR) Suppression pool Cooling......
Primary Containment Oxygen Concentration.........................
Secondary Containment..............
Secondary Containment lsolation Valves (SClvs).....
Standby Gas Treatment (SGT) System.....
PLANT SYSTEMS Residual Heat Removal service water Booster (RHRSWB) system.
Service Water (SW) System and Uttimate Heat Sink (UHSi....:..........
Reactor Equipment Cooling (REC) System.......................
Control Room Emergency Filter (CREF) System Air Ejector Offgas.......
Spent Fuel Storage PoolWater Level........
The Main Turbine Bypass System 3.4-1 3.4-1 3.4-4 3.4-6 3.4-8 3.4-10 3.4-13 3.4.8 3.4.9 3.4.10 3.5 3.5.1 3,5.2 3.5.3 3.4-15 3.4-18 3.4-20 3.4-24 3.5-1 3.5-1 3.5-7 3.5-12 3.6 3.6.1.1 3.6.1.2 3.6.1.3 3.6.1.4 3.6.1.5 3.6.1.6 3.6.1.7 3.6.1.8 3,6.1.9 3.6.2.1 3.6.2.2 3.6.2.3 3.6.3.1 3.6.4.1 3.6.4.2 3.6.4.3 3.6-1 3.6-1 3.6-3 3.6-8 3.6-16 3.6-17 3.6-18 3.6-20 3.6-23 3.6-25 3.6-27 3.6-30 3.6-31 3.6-33 3.6-34 3.6-36 3.6-40 3.7-1 3.7-1 3.7-3 3.7-6 3.7-8 3.7-11 3.7-13 3.7-14 3.7 3.7.1 3.7.2 3.7.3 3.7.4 3.7.5 3.7.6 3.7.7 (continued Cooper Amendment No.
OPS 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Overpressure protection System (OpS)
LCO 3.4.3 The OPS shall be OPERABLE APPLICABILITY: MODES 1,2, and3 ACTIONS CONDITION A. OPS inoperable.
COMPLETION TIME 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hours REQUIRED ACTION 4.1 AND 4.2 Be in MODE 3.
Be in MODE 4.
Cooper 3.4-6 Amendment No.
OPS 3.4.3 SURVEILLANCE REQUIREM ENTS sR 3.4.3.1 sR 3.4.3.2 Number of OPS SRVs Nominal Setpoint (psig) 1 080 1 090 1 100 Nominal Setpoint FREQUENCY ln accordance with the INSERVICE TESTING PROGRAM ln accordance with the INSERVICE TESTING PROGRAM SURVEILLANCE Y"ryy the as-left OPS lift pressures of the required Safety Relief Valves (SRVs) and Safety Valves (SVs) are within ! 1o/o of the nominal setpoint:
2 3
3 Number of OPS SVs (psis) 3 1240 Ygrify the as-found OPS lift pressures of the required SRVs and SVs are within the timits specified in ine COLR.
Cooper 3.4-7 Amendment No.
Reporting Requirements 5.6 5.6 Reporti ng Requirements (continued )
5.6.3 Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix l, Section lV.B.1.
5.6.4 (Deleted) 5.6.5 Core Operatinq Limits Report (COLR)
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
1 The Average Planar Linear Heat Generation Rates for specifications 3.2.1 and3.7.7.
The Minimum critical Power Ratio for Specifications 3.2.2 and 3.7.7, and M CPRgs.e"/" for Specific ation 3.2.2.
The Linear Heat Generation Rates for specifications 3.2.3 and 3.7.1.
The three Rod Block Monitor Upscale Allowabre Values for Specification 3.3.2.1.
5.
6 The Minimum critical Power Ratios in Table 3.3.2.1-1for specification 3.3.2.1.
The power/flow map defining the Stability Exclusion Region for Specification 3.4.1.
a 2
3 4
7 The as-found Overpressure Protection System lift pressures for Specification 3.4.3 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
NEDE-24011-P-A, "General Electric standard Application for Reactor Fuel" (Revision specified in the COLR).
(continued) b Cooper 1
5.Q-21 Amendment No
N1S2025035 Page 1 of 18 PROPOSED TECHNICAL SPECIFICATIONS BASES CHANGES (MARK-UP) - FOR INFORMATION ONLY Cooper Nuclear Station, Docket No. 50-298, Renewed Operating License No. DpR-46 Revised paoes ii B 2.0-6 83.1-23 B 3.3-10 B 3.3-12 B 3.3-14 B 3.3-83 B 3.3-185 B 3.4-14 through 20 B 3.4-60 B 3.6-35
TABLE OF CONTENTS B 3.3 B 3.3.7.1 B 3.3.8.1 B 3.3.8.2 INSTRUM ENTATION (continued)
Control Room Emergency Filter (CREF) System lnstrumentation..........
Loss of Power (LOP) lnstrumentation.................
Reactor Protection System (RpS) Electric Power Monitoring REACTOR COOLANT SYSTEM (RCS)
Recirculation Loops Operating..
Jet Pumps Protection System (oPS)...
RCS Operational 1EAKAGE............
RCS Leakage Detection lnstrumentation RCS Specific Activity.
Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown RCS Pressure and Temperature (p/T) Limits........
Reactor Steam Dome Pressure...
CONTAINMENT SYSTEMS.........
Primary Containment.............
Primary Containment Air Lock Primary Containment lsolation Valves (pClvs)........
Drywell Pressure Drywell Air Temperature.........
Low-Low Set (LLS) Valves Reactor Build ing{o-Suppression Cham ber Vacuum Breakers...
Suppression Chamberto-Drywell Vacuum Breakers Residual Heat Removal (RHR) Containment Spray.
Suppression Pool Average Temperature..................
Suppression Pool Water Level........
B 3.4 B 3.4.1 B 3.4.2 B 3.4.3 B 3.4.4 B 3.4.5 B 3.4.6 B 3.4.7 B 3.4.8 B 3.4.9 B 3.4.10 B 3.5 B 3.5.1 B 3.5.2 B 3.5.3 B 3.6 B 3.6.1.1 B 3.6.1.2 B 3.6.1.3 B 3.6.1.4 B 3.6.1.5 B 3.6.1.6 B 3.6.1.7 B 3.6.1.8 B 3.6.1.9 B 3.6.2.1 B 3.6.2.2 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CoNTROL, AND REACTOR CORE lSOLATlON COOLTNG (RClC)
SYSTEM...
B 3.5-1 ECCS - Operating..............
...... B 3.5_1 Reactor Pressure vessel (Rpv) water lnventory control.............8 3.5-1g RCIC System
.........8 3.5_27 B 3.3-180 B 3.3-189 B 3.3-199 B 3.4-1 B 3.4-1 B 3.4-9 B 3.4-14 B 3.4-19 B 3.4-24 B 3.4-30
.B 3.4-34
.......8 3.4-39
.......8 3.4-44
.......8 3.4-53
.... B 3.6-1
.... B 3.6-1
.... B 3.6-6
...8 3.6-15
...8 3.6-30
...B 3.6-32
...B 3.6-35
..B 3.6-39
..B 3.6-45
..8 3.6-51
..B 3.6-55
..8 3.6-60 Cooper qz4gtagxxlxxlxx I
RCS Pressure SL B 2.1.2 BASES (continued)
APPLICABLE SAFETY ANALYSES Ihe and the Reactor Protection system Reacoffissure High Function have settings established to ensure that the RCS pressure SL will not be exceeded.
The RCS pressure SL has been serected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to section llt of the ASME, Boilerand Pressure Vessel code, tgos eoition, including Addenda through the winter of 1g66 (Ref. 5), which permits a maximum pressure transient of 11oo/o,137b psig, of design pressure 1250 psig.
The sL of 1337 psig, as measured in the reactor steam dome, is equivalent to 1375 psig at the lowest elevation of the RCS. The RCS is designed to the usAS Nuclear power piping code, section B31.1,
'1967 Edition (Ref. 6), and section lll of the ASME, Boiler and pressure Vessel code, 1983 Edition (Ref. 7),for the reactor recirculation piping, which permits a maximum pressure transient of 120o/o of design pressures of 1148 psig for suction piping and 1274 psig for diicharge piping. The RCS pressure sL is serecteo to be the loriest transient overpressure allowed by the applicable codes.
SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME code, section lll, is 1 1oo/o of design piessure. The maximum transient pressure allowable in the ncs piping, valves, and fittings is 120o/o of design pressures of 114g psig for iuction piping and 1274 psig for discharge piping. The most limiting of these allowances is the 11oo/o of the RCS pressure vesseldesign piessures; therefore, the sL on maximum allowable RCS pressure is established at 1337 psig as measured at the reactor steam dome.
APPLICABILIry SL 2.1.2 appties in ail MODES.
SAFETY LIMIT VIOLATIONS Exceeding the RCS pressure sL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 cFR so.6z, "Accident source Term," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance witir tne st within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> completion Time ensures that the operators take prompt remedial action and also assures that the probability of an accident occurring during this period is minimal.
Cooper B 2.0-6 W+A#gxxlxxlxx_ l
Control Rod Scram Times B 3.1.4 BASES APPLICABLE SAFETY ANALYStS (continued)
The scram function of the cRD system protects the MCpR safety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core SLs," and LCO 3.2.i, "MINIMUM cRlrlcAl PowER MTlo (McpR)") and the 1% ctadding plastic strain fuel design limit (see Bases for LCO 3.2.1,
.AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)),
which ensure that no fuel damage will occur if these limits are not exceeded. Above 800 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCpR from becoming less than the MCPR sL, du-ring the analyzed limiting power transient. Below 800 psig, the scram function is assumed to pLrform during the control rod drop accident (Ref. s) and, therefore, aiso provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCo 3.1.6, "Rod pattern control"). For the reactor vessel overpressure protection analysis, the scram function, along with the sa+e+#re+ie+rya+*esoverpressure protection Svstem, ensure that the peak vessel pressure is maintained within the applicable ASME code limits.
control rod scram times satisfy criterion 3 of 10 cFR 50.36(c)(2)(ii) (Ref 6).
LCO The scram times specified in Table 3.1.4-1 are required to ensure that the ggral reactivity assumed in the DBA and transient analysis is met (Ref.
7). To account for single failures and "slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis. The scram times have a margin that allows up to approximately 7o/o of the control rods (e.g., 137 x7o/o
- 10) to have scram times exceeding the specified limits (i.e., "slow" control rods) assuming a single stuck control rod (as allowed by LCo 3.1.3, "control Rod OPERABILITY') and an additional control rod failing to scram per the single failure criterion. The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes ("pickup") when the index tube passes a specific location and then opens ("dropout") as the index tube travels upward. Verification of the specified scram times in Table 3.1.4-1 is (continued)
Cooper B 3.1-23 nevisien+xlXlXl I
RPS lnstrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALysls, Lco, and AppLtcABtLtry (continued) 2.c. Averaqe Po*er Ranoe Monitor Neutron Flux-Hioh (Fixed)
The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases.
The Average Power Range Monitor Neutron Flux-High (Fixed) Function is capable of generating a trip signal to prevent fuel damage or excessive Reactor coolant system (Rcs) pressure. For the overfressurization protection analysis of Reference 7, the Average power Range Monitor Neutron Flux-High (Fixed) Function is assumed to terminatsthe main steam isolation valve (MSIV) closure event and, along with the _
, limits the peak reactor pressure vessel (Rpv) pressure to less inan tne ASME code limits. The control rod drop accident (CRDA) anatysis (Ref. g) takes credit for the Average Power Range Monitor Neution Fiux-High (Fixed) Function to terminate the CRDA.
The APRM system is divided into two groups of channels with three APRM channels inputting to each trip system. The system is designed to allow one channel in each trip system to be bypassed. Any one ApRM channel in a trip system can cause the associated trip system to trip.
Four channels of Average Power Range Monitor Neutron Flux-High (Fixed) with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. ln addition, to provide adequate coverage of the entire core, at least 11 LPRM inputs are required for each APRM channel, with at least two LPRM inputs from each of the four axiar revels at which the LpRMs are located.
The Allowable Value is based on the Anaryticar Limit assumed in the CRDA analyses.
The Average Power Range Monitor Neutron Flux-High (Fixed) Function is required to be OPERABLE in MODE 'l where the potential consequences of the analyzed transients could result in the sls (e.g., McpR and RCS pressure) being exceeded. Although the Average power Range Monitor FrJtgl Flux-High (Fixed) Function is assumed in the CRDA Jna[sis (Ref. 8), which is applicable in MODE 2, the Average power Range Monitor Neutron Flux-High, (startup) Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average power Range Monitor Neutron Flux-High (Fixed) Function is not required ln UODE Z.
Cooper B 3.3-10 11.1%l-l2xxlxxlxx
RPS lnstrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALysls, Lco, and AppLlcABtLtry (continued) one APRM in each trip system may be bypassed, only one APRM in each trip system may be inoperable without resulting in an ilps trip signal.
This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the Rps as required by the NRC approved licensing basis.
Four channels of Average power Range Monitor-lnop with two channels in each trip system are required to be bpgnnglE to ensure that no single failure will preclude a scram from this Function on a valid signal.
There is no Allowable Value for this Function.
This Function is required to be opERABLE in the MODES where the APRM Functions are required.
- 3.
Reactor Vessel pressure-Hiqh An increase in the RpV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL powER transfeired to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysii tikes direct credit for this Function. However, the Reactor Vessel preisure-High Function initiates a scram for transients that result in a pressure inirease, counleracting the pressure increase by rapidly reducing core power.
For the overpressurization protection inalysis of Refer6nce 7, reactor lcram (the analyses conservatively assume scram on the Average power Range Monitor Neutron Flux-High (Fixed) signal, not the Reactor Vessel Pressure-High signat), tong with the SRte@
svsteryr, limits the peak RpV pressure to tess thanlhe ASME Section ttt Code limits.
High reactor pressure signals are initiated from four pressure switches tfgt sense reactor pressure. The Reactor Vessel pressure-High Allowable Value is chosen to provide a sufficient margin to the ASME Section lll Code limits during the event.
Four channels of Reactor Vessel pressure-High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be oPEMBLE to ensure that no single instrument failure will preclude a scram from this Function on a valid srgnal. The Function is required to be oPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.
Cooper B 3.3-12
+aL5l-12xxtlx>nlxx I
RPS lnstrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALysts, Lco, and AppLtcABtLtry (continued) the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization transient.
However, for the overpressurization protection analysis of Reference 7, the Average Power Range Monitor Neutron Flux-High (Fixed) Function, along with the sR\\boverpressure protection systeir, iimits the peak RPV pressure to less than the ASME cooe timits. rnat is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis.
The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Eccs, ensures that the fuel peak cladding temperature remains below the limits of 10 cFR 50.46.
MSIV closure signals are initiated from position switches located on each gtl!" eight MSlVs. Each MSlv has two position switches; one inputs to RPS trip system A while the other inputs to Rps trip system B. Each RPS trip system receives an input from four trrtain dteim lsolation Valve-closure channels, each consisting of two position switches (one for the inboard MSlv and one for the outboird MSlv in the same steam line) in series with a sensor relay. The logic for the Main Steam lsolation Valve-closure Function is arranged iuch that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to.occur. The design permits closure of any two lines without a full scram being initiated.
The Main Steam lsolation Valve-closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subiequent pressure transient.
Eight channels of the Main steam lsolation Valve-closure Function, with four channels in each trip system, are required to be OpERABLE to ensure that no single instrument failure will preclude the scram from this Function on a valid signal. This Function is only required in MoDE 1 since, with the MSIVs open and the heat generition rate high, a pressurization transient can occur if the lvlSlvs close. tn uooe 2, the heat generation rate is low enough so that the other diverse RpS functions provide sufficient protection.
6 Drvwell Press oh High pressure in the drywellcould indicate a break in the RcpB. A reactor scram is initiated to minimize the possibility of fuel damage and to reduce the amount of energy being added to the coolant and thJdrywell.
The Drywell Pressure-High Function is assumed to scram the reactor Cooper B 3.3-14 111?51+2xxlxxlxx I
ATWS-RPT I nstru mentation B 3.3.4.1 BASES APPLICABLE sAFEry ANALYSIS, Lco, and AppLtcABtLtry (continued)
- b.
Reactor Pressure-Hiqh Excessively high RPV pressure may rupture the RCpB. An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity
.insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization.
The Reactor Pressure-High Function initiates an RpT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the sa+e+V#e+ief vatvesOverpressure protec
, limits the peak RpV pressure to less than the ASME Section lll limits.
The Reactor Pressure-High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure.
Four channels of Reactor Pressure-High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor pressure-High Allowable Value is chosen to provide an adequate margin to the ASME Section lll limits.
ACTIONS A Note has been provided to modify the ACTToNS related to ATWS-Rpr instrumentation channels. section 1.3, completion Times, specifies that once a condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the condition, discovered to be inoperable or not within limits, will not resurt in separate entry into the condition. section 1.3 also specifies that Required Actions of the condition continue to apply for each additionar failure, with completion Times based on initial entry into the condition. However, the Required Actions for inoperable ATWS-Rpr instrumentation channels provide appropriate compensatory measures for separate inoperable channels.
As such, a Note has been provided that ailows separate condition entry for each inoperable ATWS-RPT instrumentation channel.
A.1 and A.2 with one or more channels inoperable, but with ATWS-RPr capability for each Function maintained (referto Required Actions B.1 and c.1 Baies),
the ATWS-RPT System is capable of performing the intended function.
However, the reliability and redundancy of the ATWS-RPT Cooper B 3.3-83 111%tl2x4lxxlxx I
LLS lnstrumentation B 3.3.6.3 BASES SURVEI LLANCE REQU I REMENTS (continued) sR 3.3.6.3.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required actuation logic for a specified channer.
The system functional testing performed in tCO 3,4,3; "safety/Relref LCO 3.0.1.6, "Low-Low Set (LLS) safety/Relief Valves (sRVs)," for sRVs overlaps this test to provide complete testing of the assumed safety function.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1.
USAR, Section tV-4.5.2 2.
3.
1 0 cFR 50.36(c)(2)(ii).
GENE-770-06-1, "Bases for Changes to Surveillance Test lntervals and Allowed Out-of-Service Times for Selected lnstrumentation Technical Specifications," February 1 gg1.
Cooper B 3.3-185 Wxxlxxlxx
B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.3 (oPS)
BASES re BACKGROUND The Protection OPS nts ove ressuri reactor steam. This action rotects tn e reactor failure ch could result in the release of fission ucts Ref. 1 The sRVs and sVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. Each of the eioht SRVs discharq es steam throuqh a discharoe Iine to a ooint below the minimum water level in the su ressron The three steam rectl to the the relief mode or actuated mode of n
a remote controlled umatic is fitted ot ssem selective ration of the valve at ures from to valve set re IS isad m
c ctuator must be actuated to n
valve actuated eans of a d
valve to the chamber and strokes umati ste in turn suoolied to accomplish this action. Deeneroizi no the solenoid The sRVs can actuate by either of two modes: the safety mode or the relief mode. However, for the purposes of this LCo, only the safety mode is required. ln the safety mode (or spring mode of operaiion), the siring loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed. opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve. This satisfies the code requirement.
The sVs only operate in the safety mode. They are spring-loaded and were designed and constructed in accordance with tne nSue Boiler and Pressure Vessel, section lll, Article 9, code requirements. The salgly e
r ure Cooper manuallv B 3.4-14 Revisren4x/x/>q I
B 3.4.3 ue atic disc to the set ure sorino oreload. The one c ooerator is a so that its malfunction will not orevent valve disk from liftinq if steam inlet reaches the s lift set ln the relief mode valves be manual ora at th ress overpressure protection, relief in the for LCO 3.3.6.3 "Low-Low Set instru 3.3.5.1 "Em n
Core lnstrumentation."
for the s ECCS APPLICABLE SAFETY ANALYSES The everpressure preteetien systenrops. must accommodate the most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSlVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. zlJ. A+epa+ate Cooper B 3.4-15 Revisien Oxx/xx/xx
SRV-andVsOPS B 3.4.3 BASES APPLICABLE SAFETY ANALYSES (continued)
@The disch is desi to accommodate forces resuItinq from the relief action includino intera ns with the su ron ool and is su for reactions due to flow at maximum SRV di ove re on Ref e
safetv mode of operation. aAnalysis results demonstrate that the desgn SnV anA SV eapaeit,,
gpS, is of maintaining reactor pressure below the ASME code limit of 1 capable 10o/o of vessel design pressure (110Vo x 1250 psig = 1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is mef during the mest severe-desi g n bas is preseu+e-*ransientevent.
SRVs and SVsIheAPS sati+satisfies Criterion 3 of 10 CFR
- 50. 36(c)(2)(i i )'{Pst+).
From an overpressure standpoint, the design basis events are bounded by the MSlv closure with flux scram event described above. Reference,#
.,ldiscusses additional events that are expected to actuate the sRVs and SVs.
LCO The is r+*r++e$e OPERABLE when it can ensure that the ASM E Code limit on k
reactor ressure. as tn Safetv Limit2
- 2. will be orotected usinq e
eof and meehenieafiyepen-to relieve excess pressure and maintain reactor below Limit2.1.2 and credit less than the full lement lled S nd (safety funtien)
The sRV and sV setpoints are established to ensure that the ASME code limit on peak reactor pressure is satisfied. The ASME code specifications require the lowest safety valve setpeinlto be set at or below vessel design pressure (12s0 psig) and the highest sEty valve to be set so that the total accumulated pressure does not exceed 110o/o of the design pressure for conditions. The transient evaluations in Reference 3 are based on these setpoints, but also include the-additional uncertainties to provide an added degree of conservatism.
Cooper B 3.4-16 SSl05ll2xxlxxlxx
W B 3.4.3 BASES APPLICABLE SAFEry ANALYSES (continued) could result in a more severe reactor response to a transient than predicted, possibly resulting in the Safety Limit 2.f.i. being eiceeded.
Cooper B 3.4-17 e3#F/azrarlxxlyx I
W B 3.4.3 BASES APPLICABILITY ln MODES 1,2, and 3, 7 ef I SRVs and B SVsthe OpS must be OPERABLE-since there mav be considerable energy pnay$e-in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The sRVs and sVsops may be required to provide pressure relief to limit peak reactor pressure.
ln MoDE 4, decay heat is low enough for the RHR system to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents. ln MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The sRV and sv funtienOPg is not needed during these conditions.
ACTIONS 4.1 a 4.2 with the safety funetien ef ene er mere ef the required sRVs er svs
, a transient may result in the violation of the ASME Code limit on reactor pressure. lf+ndW er-mere.f+he-reguired sRVs er svs is ineperable, tfhe plant must be brought to a MoDE in which the LCo does not apply. To'achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE REQU IREM ENTS sR 3.4.3.1 This Surveillance pressures assumed in the safety analysis ef Referenee 3, The demenstratien ef the sRV and sV safety funetien lift settings must be
,e_eene_in a"rdene-w+thveri f i es that the SVs and SRVs bv the OPS have their as-left setti within 1o/o of the nominal onenrno oressrrre nos nt. The OPS m credit less than the full com ent of instal SVs a and the on a
ies to those and SRVs ired to the L on in accordance with the INSE RVICE TESTING PROGRAM correspond to ambient temperatures and pressures. The SVs and SRVs hewever, the valve+are reset to t 1% during the surveillance to allow for drift.
Cooper B 3.4-18 Aglggqxxtxx/L I
W B 3.4.3 BASES SURVEILLANCE REQUIREMENTS (continued)
SR3.4.3.2 with the reeemmendatiens ef the vender, \\dequate steam flew is flew+406 lb/hr, Plant startup is allewed prier te perferming this test preteetien are verified, per ASME Gede requirements, prier te valve are reaehed is suffieientte aehieve stable eenditiens fer testing and tift res of the SVs and with the assum ns of the ure ana is. The OPS m credit less than e
co lement lled onl a
VA r
The ent of the SVs and SRVs lift ressures must be ed in accordance with the INSE OPS and as found lift ressures are s to RVs to meet the not be for this ma Cooper B 3.4-19 tn the COLR.
g&lll+l7lxxlxxlxx I
W B 3.4.3 BASES REFERENCES 1,
ASME Beiler. and Pressure Vessel Gede, Seetien lll, 2L USAR, Section lV-4.9
- 2.
ASME Boiler and Pressure Vessel Code, Section lll.
3.
. USAR, Section XlV.
4 USAR; Seetien Xl\\A
- Plants, Cooper B 3.4-20 g}rcS42xxlxxlxx.
I
Reactor Steam Dome Pressure B 3.4.10 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.10 Reactor Steam Dome pressure BASES BACKGROUND The reactor steam dome pressure is an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria and is also an assumed initial condition of design basis accidents and transients.
APPLICABLE SAFETY ANALYSES The reactor steam dome pressure of = 1o2o psig is an initial condition of the vessel overpressure protection analysis oiReference 1. This analysis assumes an initial maximum reacior steam dome pressure and evaluates the respo_nse of the @;r,ty+he saa+#relie++a*esove rp ress u Le protectio n svite m.; d u ri n g tn"e t i m iti n g pressurization transient. The oetermination of cornpliance-with tne overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conseryed.
Reference 2 also assumes an initial reactor steam dome pressure for the analyses of design basis accidents and transients used to determine the limits for fuel cladding integrity (see Bases for LCo 3.2.2,,MlNlMUM cRlrlcAl PowER RATIO (McpR)") and 1o/o fuel cladding prastic strain (see Bases for LCO 3.2.1, "AVERAGE PLANAR LTNEAR ilfnf GENERATTON RATE (APLHGR)").
Reactor steam dome pressure satisfies the requirements of criterion 2 of 10 cFR 50.36(c)(2)(ii) (Ref. 3).
LCO The specified reactor steam dome pressure limit of < 1o2o psig ensures the plant is operated within the assumptions of the reactor overpressure protection analysis. operation above the limit may result in a response more severe than analyzed.
APPLICABILITY ln MoDES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. ln these Cooper B 3.4-60 nevisien-x/xx/xx.
I
LLS Valves B 3.6.1.6 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.6 Low-Low Set (LLS) Vatves BASES BACKGROUND The safety/relief valves (sRVs) can actuate in either the safety mode as part of the Overpressure Protectiplr System, the Automatic Depressurization system mode, or the LLS mode. ln the LLS mode (or power actuated mode of operation), a pneumatic diaphragm and stem assembly overcomes the spring force and opens the piloivalve. As in the safety mode, opening the pilot valve allows a differential pressure to develop across the main valve piston and opens the main valve. The main valve can stay open with valve inlet steam pressure as low as 50 psig. Below this pressure, steam pressure may not be sufficient to hold the main valve open against the spring force of the pilot valves. The pneumatic operator is arranged so that its malfunction will not prevent the valve disk from lifting if steam inlet pressure exceeds the safety mode pressure setpoints.
Two of the sRVs are equipped to provide the LLS function. The LLS logic causes the LLS valves to be opened at a lower pressure than the relief or safety mode pressure setpoints and stay open longer, so that reopening more than one sRV is prevented on subsequeni actuations.
Therefore, the LLS function prevents excessive short duration SRV cycles with valve actuation at the relief setpoint.
Each sRV discharges steam through a discharge line and quencher to a location near the bottom of the suppression pool, which causes a load on the suppression poolwall. Actuation at lower reactor pressure results in a lower load.
APPLICABLE SAFETY ANALYSES The LLS relief mode functions to ensure that the containment design basis of one sRV operating on "subsequent actuations" is met. ln-other words, multiple simultaneous openings of sRVs (following the initial opening), and the corresponding higher loads, are avoidel. The safety analysis demonstrates that the LLS functions to avoid the induced thrust loads on the sRV discharge line resulting from "subsequent actuations" of the sRV during Design Basis Accidents (DBAs). Furthermore, the LLS function justifies the primary containment analysis assumption that simultaneous SRV (continued)
Cooper B 3.6-35 Revisien 0xxlxxlxx
N152025035 Page 1 of 2 EXAMPLE UPDATED CORE OPERATING LIMITS REPORT - FOR INFORMATION ONLY Cooper Nuclear Station, Docket No. 50-298, Renewed Operating License No. DpR-46
- 5. OVERPRESSURE PROTECTION SYSTEM 5.1 Technical Specification Reference Technical Specifi cation 3.4.3.
5.2 As-Found Lift Pressures The as-found overpressure Protection system (ops) rift pressures of the required Safety/Relief Valves (SRV) and Safety Valves (SV) shall be within the limits defined in Table 5-1.
Table 5-1: As-Found OPS Lift Pressure Limits Valve Type Number of Valves As-Found OPS Lift Pressure Limit (psiq)
SRV 1
31112.4 SRV 3
<1122.7 SRV 3
<1133.0 SV 3
<1277.2