NLS2019011, License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF-514, Revision 3, Revise BWR Operability Requirements and Actions for RCS Leakage Instrumentation.

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License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF-514, Revision 3, Revise BWR Operability Requirements and Actions for RCS Leakage Instrumentation.
ML19071A111
Person / Time
Site: Cooper Entergy icon.png
Issue date: 02/28/2019
From: Dent J
Nebraska Public Power District (NPPD)
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2019011
Download: ML19071A111 (38)
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Text

H Nebraska Public Power District Always there when you need us 50.90 NLS2019011 February 28, 2019 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

License Amendment Request for Adoption of Technical Specification Task Force Traveler TSTF-514, Revision 3, "Revise BWR Operability Requirements and Actions for RCS Leakage Instrumentation" Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46

Dear Sir or Madam:

In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CPR), Nebraska Public Power District (NPPD) is submitting a request for an amendment to the Technical Specifications (TS) for Cooper Nuclear Station (CNS).

The proposed amendment would revise the TS to define a new time limit for restoring inop~rable Reactor Coolant System (RCS) leakage detection instrumentation to operable status; establish alternate methods of monitoring RCS leakage when one or more required monitors are inoperable; and make TS Bases changes which reflect the proposed changes and more accurately reflect the contents of the facility design basis related to operability of the RCS leakage detection instrumentation. These changes are consistent with the NRC-approved Revision 3 to Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-514, "Revise BWR Operability Requirements and Actions for RCS Leakage Instrumentation." The. availability of this TS improvement was announced in the Federal Register on December 17, 2010 (75 FR 79048), as part of the consolidated line item improvement process.

  • Attachment 1 provides an evaluation of the proposed changes.
  • Attachment 2 provides the markup pages of existing TS to show the proposed changes.
  • Attachment 3 provides the markup pages of the existing TS Bases to show the proposed changes.
  • Attachment 4 provides revised (clean) TS pages.

NPPD requests approval of the proposed license amendment by February 28, 2020, with the amendment being implemented within 60 days.

~OOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com

NLS2019011 Page 2 of3 The proposed TS changes have been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review and Audit Board). 'Amendments to the CNS Facility Operating License through Amendment 262 dated February 21, 2019, have been incorporated into this request. This request is submitted under oath in accordance with 10 CFR 50.30(b).

In accordance with 10 CFR 50.9l(a)(l), "Notice for Public Comment," the analysis regarding the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is being provided to the Commission in accordance with the distribution requirements in 10 CFR 50.4.

In accordance with 10 CFR 50.91(b)(l), "State Consultation," a copy of this application and its reasoned analysis about no significant hazards considerations is being provided to the designated State. of Nebraska Official.

This letter contains no regulatory commitments.

If you should have any questions about this submittal please contact Jim Shaw, Licensing Manager, at (402) 825-2788.

I declare under penalty of perjury that the foregoing is true and correct.

Executed On: [). f:J t-f J.o f9 Date Sincerely, Vice President - Nuclear and Chief Nuclear Officer

/tf Attachments: 1. Description and Assessment of Technical Specifications Changes

2. Proposed Technical Specifications Changes (Mark-up)
3. Proposed Technical Specifications Bases Changes (Mark-up) -

Information Only

4. Revised Technical Specifications Pages cc: Regional Administrator w/ attachments USNRC - Region IV Cooper Project Manager w/ attachments USNRC - NRR Plant Licensing Branch IV

NLS2019011 Page 3 of3 Senior Resident Inspector w/ attachments USNRC-CNS Nebraska Health and Human Services w/ attachments Department of Regulation and Li censure NPG Distribution w/o attachments CNS Records w/ attachments

NLS2019011 Page 1 of7 Attachment 1 Description and Assessment of Technical Specifications Changes Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 1.0 Description 2.0 Proposed Changes 3.0 Background 4.0 Technical Analysis 5.0 Regulatory Analysis 5.1 No Significant Hazards Consideration Analysis 5 .2 Applicable Regulatory Requirements/Criteria 6.0 Environmental Consideration

7. 0 References

NLS2019011 Page 2 of7

1.0 DESCRIPTION

The proposed amendment would revise the Technical Specifications (TS) to define a new time limit for restoring inoperable Reactor Coolant System (RCS) leakage detection instrumentation to operable status; establish alternate methods of monitoring RCS leakage when one or more required monitors are inoperable; and make conforming TS Bases changes. These changes are consistent with the NRC-approved Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-514, Revision 3, "Revise BWR Operability Requirements and Actions for RCS Leakage Instrumentation." The availability of this TS improvement was announced in the Federal Register on December 17, 2010 (75 FR 79048), as part of the consolidated line item improvement process (CLIIP).

2.0 PROPOSED CHANGE

S The proposed changes revise and add a new Condition C to TS 3.4.5, "RCS Leakage Detection Instrumentation," and revise the associated bases. New Condition C is applicable when the drywell atmosphere gaseous radiation monitor is the only operable TS-required instrument monitoring RCS leakage, i.e., TS-required particulate and sump monitors are inoperable. New Condition C Required Actions require monitoring RCS leakage by obtaining and analyzing grab samples of the primary containment atmosphere every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; monitoring RCS leakage using administrative means every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; and taking action to restore monitoring capability using another monitor within 7 days.

Additionally, the TS Bases which summarize the reasons for the specifications, are revised to clarify the specified safety function for each required instrument in Limiting Condition for Operation (LCO) Bases, delete discussion from the TS Bases that could be construed to alter the meaning of TS operability requirements, and reflect the changes made to TS 3.4.5.

Nebraska Public Power District (NPPD) is proposing variations or deviations from the TS changes described in TSTF-514, Revision 3, and the NRC staffs model Safety Evaluation (SE) published in the Federal Register on December 17, 2010 (75 FR 79048), as part of the CLIIP Notice of Availability.

The change being proposed by NPPD affects the TS Bases changes associated with TSTF-514, Revision 3. The TS Bases markup pages attached to TSTF-514 contains a statement regarding the gaseous portion of the drywell atmospheric radioactivity monitor in the LCO section that is not supported by Cooper Nuclear Station's (CNS) design basis calculations. This statement is as follows:

"However, the gaseous or particulate drywell atmospheric radioactivity monitor is OPERABLE when it is capable of detecting a 1 gpm increase in unidentified LEAKAGE within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> given an RCS activity equivalent to that assumed in the design calculations for the monitors."

NLS201901 l Page 3 of7 This statement is being revised to the following:

"However, the particulate drywell atmospheric radioactivity monitor is OPERABLE when it is capable of detecting a 1 gpm increase in unidentified LEAKAGE within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> given an RCS activity equivalent to that assumed in the design calculations for the monitors.

Additionally, the gaseous drywell atmospheric radioactivity monitor is OPERABLE when it is capable of detecting a LEAKAGE rate ofless than the LEAKAGE rate limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> given an RCS activity equivalent to that assumed in the design calculations for the monitors."

The background and basis for this change will be discussed in the Technical Analysis and Regulatory Analysis sections to follow. This difference does not alter the conclusion that the

  • proposed change is applicable to CNS.

3.0 BACKGROUND

NRC Information Notice (IN) 2005-24, "Nonconservatism in Leakage Detection Sensitivity,"

dated August 3, 2005, informed addressees that the reactor coolant activity assumptions for primary containment atmosphere gaseous radioactivity monitors may be non-conservative. This means the monitors may not be able to detect a one gallon per minute (gpm) leak within one hour. Some licensees, in response to IN 2005-24, have taken action to remove the gaseous radiation monitor from the TS list of required monitors. However, industry experience has shown that the primary containment atmosphere gaseous radiation monitor is often the first monitor to indicate an increase in RCS leak rate. As a result, the TSTF and the NRC staff met on April 29, 2008, and April 14, 2009, to develop an alternative approach to address the issue identified in IN 2005-24. The agreed solution is to retain the primary containment atmosphere gaseous radiation monitor in the LCO list of required equipment, revise the specified safety function of the gas monitor to specify the required instrument sensitivity level, to revise the Actions requiring additional monitoring, and provide less time before a plant shutdown is required when the primary containment atmosphere gaseous radioactivity monitor is the only operable monitor.

4.0 TECHNICAL ANALYSIS

NPPD has reviewed TSTF-514, Revision 3, and the model SE published in the Federal Register on December 17, 2010 (75 FR 79048), as part of the CLIIP Notice of Availability. NPPD has concluded that the technical bases presented in TSTF Traveler-514, Revision 3, and the model SE prepared by the NRC staff are applicable to CNS.

The TSTF-514 Traveler and SE discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A{ General Design Criteria (GDC). CNS was not licensed to the 10 CFR 50, Appendix A, GDC. CNS was designed and constructed to meet the principle design criteria described in the Atomic Energy Commission's (ABC) proposed rule, "General Design Criteria for Nuclear Power Plant Construction Permits," published in the Federal Register on July 11, 1967 (32 FR 10213 ). The degree of conformance to the 1967 proposed GDC is described in Appendix F, "Conformance to AEC Proposed General Design Criteria" to the

NLS2019011 Attachment 1 Page 4 of7 Updated Safety Analysis Report (USAR) for CNS. This difference does not alter the conclusion that the proposed change is applicable to CNS.

CNS' current licensing basis incorporates the proposed GDC that are equivalent to the 10 CFR Part 50, Appendix A, GDC 30. The proposed amendment is consistent with the AEC proposed GDC in that the design requirements for instrumentation, reactor coolant leak detection, the reactor coolant pressure boundary, and reactor coolant makeup are unaffected. Regulatory Guide (RG) 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973, describes acceptable methods of implementing this requirement with regard to the selection of leakage detection systems for the reactor coolant boundary. CNS is not committed to RG 1.45; however, with the exception of the gaseous channel sensitivity CNS conforms to the regulatory position outlined in RG 1.45. Section IV-10, "Reactor Coolant System Leakage Rate Limits," of the CNS USAR provides details associated with the containment atmospheric leakage detection systems in use at CNS.

A review of CNS' licensing and design basis determined that the design of the gaseous portion of the drywell atmospheric radioactivity monitor could not meet the sensitivity requirements outlined in RG 1.45 for the reasons discussed in TSTF-514, Revision 3. Due to the absence of a commitment in the CNS licensing basis to RG 1.45, a design basis coolant activity corresponding to the sensitivity of the gaseous portion of the drywell atmospheric radioactivity monitor had not been previously developed. With this discovery, rather than assuming an unrealistic RCS coolant activity in the design calculations for the gaseous portion of the drywell atmospheric radioactivity monitor that would provide the ability to detect a one gpm leak in the RCS boundary within one hour, the current in situ values for RCS coolant activity were used. Using these values, it was determined that the gaseous portion of the drywell atmospheric radioactivity monitor could detect a leakage rate ofless than the leakage rate limits as defined by CNS Technical Specifications, within one hour. This information was then included in the design basis calculations for the gaseous portion of the drywell atmospheric radioactivity monitor to support the proposed changes to the TS Bases associated with this License Amendment Request.

This difference does not alter the conclusion that the proposed change is applicable to CNS.

The administrative means of monitoring include diverse alternative mechanisms from which appropriate indicators may be selected based on plant conditions. NPPD will utilize the following method or methods considering the current plant conditions and historical or expected sources of unidentified leakage: drywell equipment sump temperature, suppression pool water level, primary containment pressure, and primary containment temperature.

There are diverse alternative methods for determining that RCS leakage has not increased, from which appropriate indicators may be selected based on plant conditions. NPPD will utilize the following method or methods considering the current plant conditions and historical or expected .

sources of unidentified leakage: drywell equipment sump temperature, suppression pool water level, primary containment pressure, and primary containment temperature. Actions to verify that these indications have not increased since the required monitors became inoperable and analyze primary containment atmospheric grab samples are sufficient to alert the operating staff to an unexpected increase in RCS leakage.

NLS201901 l Page 5 of7

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Analysis Nebraska Public Power District (NPPD) has evaluated the proposed changed to the Technical Specifications (TS) using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration. An analysis of the issue ofno significant hazards consideration is presented below:

Description of Amendment Request: The proposed amendment would revise TS 3.4.5, "Reactor Coolant System (RCS) Leakage Detection Instrumentation," Conditions and Required Actions and the licensing basis for the drywell atmospheric particulate and gaseous radiation monitors, as well as make associated TS Bases changes for TS 3.4.5.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the NPPD analysis of the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is presented below: *

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the drywell atmospheric gaseous radiation monitor. The monitoring of RCS leakage is not a precursor to any accident previously evaluated. The monitoring of RCS leakage is not used to mitigate the consequences of any accident previously evaluated. Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the drywell atmospheric gaseous radiation monitor. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

NLS2019011 Page 6 of7

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change clarifies the operability requirements for the RCS leakage detection instrumentation and reduces the time allowed for the plant to operate when the only TS-required operable RCS leakage detection instrumentation monitor is the drywell atmospheric gaseous radiation monitor. Reducing the amount of time the plant is allowed to operate with only the drywell atmospheric gaseous radiation monitor operable increases the margin of safety by increasing the likelihood that an increase in RCS leakage will be detected before it potentially results in gross failure. Therefore, it is concluded that this change does not involve a significant reduction in a margin of safety.

Based on the above analysis, NPPD concludes that the requested change does not involve a significant hazards consideration as set forth in 10 CFR 50.92(c), "Issuance of Amendment."

5.2 Applicable Regulatory Requirements/Criteria A description of the proposed TS change and its relationship to applicable regulatory requirements were published in the Federal Register Notice of Availability on December 17, 2010 (75 FR 79048). NPPD has reviewed the NRC staffs model SE referenced in the CLIIP Notice of Availability and concluded that the regulatory evaluation section is not applicable to CNS. The following regulatory requirements apply to CNS.

CNS was designed and constructed to meet the principle design criteria described in the AEC's proposed rule, "General Design Criteria for Nuclear Power Plant Construction Permits,"

published in the Federal Register on July 11, 1967 (32 FR 10213). The degree of conformance to the 1967 proposed GDC is described in Appendix F, "Conformance to AEC Proposed General Design Criteria" to the USAR for CNS. CNS' current licensing basis incorporates the proposed GDC that are equivalent to the 10 CFR Part 50, Appendix A, GDC 30. Criterion 30, "Quality of reactor coolant pressure boundary," requires that means be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage. The proposed license amendment is consistent with the AEC proposed GDC in that the design requirements for instrumentation, !eactor coolant leak detection, the reactor coolant pressure boundary, and reactor coolant makeup are unaffected.

RG 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973, describes acceptable methods of implementing this requirement with regard to the selection of leakage detection systems for the reactor coolant boundary. The position ofRG 1.45 is that at least three different detection methods should be employed. Two of these methods should be: (1) sump level and flow monitoring and (2) airborne particulate radioactivity monitoring. The third method may involve either monitoring, of condensate flow rate from air coolers or monitoring of gaseous radioactivity. The RG recommends that the sensitivity and response time of each leakage detection system employed for unidentified leakage should be adequate to detect a

NLS2019011 Page 7 of7 leakage rate, or its equivalent, of one gpm in less than one hour. CNS conforms to the regulatory position outlined in RG 1.45. However, as discussed in the Technical Analysis, the gaseous portion of the drywell atmospheric radiation monitor does not meet the sensitivity requirements ofRG 1.45 nor is CNS specifically committed to RG 1.45. This difference does not alter the conclusion that the proposed change is applicable to CNS.

Section IV-10, "Reactor Coolant System Leakage Rate Limits," of the CNS USAR provides details associated with the containment atmospheric leakage detection systems in use at CNS.

TS 3 .4.5 establishes LCOs for three of these systems: (1) the drywell floor drain sump flow monitoring system, (2) the drywell atmospheric particulate monitoring system, and (3) the drywell atmospheric gaseous monitoring system. As discussed in the USAR, drywell equipment sump temperature, suppression pool water level, primary containment pressure, and primary containment temperature also provide a means for detecting leaks within the primary containment.

6.0 ENVIRONMENTAL CONSIDERATION

The proposed change 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 change does not involve (i) a*

significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22( c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

7 .0 REFERENCES

l. Federal Register Notice of Availability published on December 17, 2010 (75 FR 79048).
2. TSTF-514-A, Revision 3, "Revise BWR Operability Requirements and Actions for RCS Leakage Instrumentation."
3. Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems,"

Revision 0, May 1973.

NLS2019011 Page 1 of 4 Attachment 2 Proposed Technical Specifications Changes (Mark-up)

Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 TOCpageii 3.4-11 3 .4-11 Insert

TABLE OF CONTENTS (continued) 3.4 REACTOR COOLANT SYSTEM (RCS) ....................................................... 3.4-1 3.4.1 Recirculation Loops Operating ............................................................... 3.4-1 3.4.2 Jet Pumps .............................................................................................. 3.4-4 3.4.3 Safety/Relief Valves (SRVs) and Safety Valves (SVs) ........................... 3.4-6 3.4.4 RCS Operational LE~KAGE .............._. ................................................... 3 . 4 - 8 ~

3.4.5 RCS Leakage Detection Instrumentation ............................................... 3.4-10 3 3.4.6 RCS Specific Activity ............................................................................. 3.4-3.4.7 Residual Heat Removal (RHR) Shutdown Cooling ~

System - Hot Shutdown .................................................................... 3.4-44 Res:Y~~~~~ai::,~~~:~~~~~~.~.~.~.t~~~~-~~~~'.~~

3.4.8

................................. 3. 4 - ~

3.4.9 RCS Pressure and Temperature (PIT) Limits .........................................3 . 4 - ~

3.4.10 Reactor Steam Dome Pressure ............................................................. 3 . 4 - 2 3 - ~

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPVWATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION I

COOLING (RCIC} SYSTEM ........................................................................ 3.5-1 3.5.1 ECCS - Operating .................................................................................. 3.5-1 3.5.2 RPV Water Inventory Control .................................................................3.5-7 3.5.3 RCIC System ............._............................................................................3.5-12 3.6 CONTAINMENT SYSTEMS ......................................................................... 3.6-1 3.6.1.1 Primary Containment ............................................................:., ................ 3.6-1 3.6.1.2 Primary Containment Air Lock................................................................ 3.6-3 3.6.1.3 Primary Containment Isolation Valves (PCIVs) ......................................-3.6-8 3.6.1.4 Drywell Pressure .....................................................:............. ,................ 3.6-16 3.6.1.5 Drywell Air Temperature ........................................................................ 3.6-17 3.6.1.6 Low-Low Set (LLS) Valves ..................................................................... 3.6-18 3.6.1.7 Reactor Building-to-Suppression Chamber Vacuum Breakers ............... 3.6-20 3.6.1.8 Suppression-Chamber-to-Orywell Vacuum Breakers ............................. 3.6-23 3.6.1.9 Residual Heat Removal (RHR) Containment Spray ............................... 3.6-25 3.6.2.1 Suppression Pool Average Temperature ............................................... 3.6-27 3.6.2.2 Suppression Pool Water Level ............................................................... 3.6-30 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling ..................... 3.6-31 3.6.3.1 . Primary Containment Oxygen Concentration .......................................... 3.6-33 3.6.4.1 Secondary Containment ......................................................................... 3.6-34 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) .................................. 3.6-36 3.6.4.3 Standby Gas Treatment (SGT) System .................................................. 3.6-40 3.7 PLANT SYSTEMS .....................................................................................,.3.7-1 3.7.1 Residual Heat Removal Service Water Booster (RHRSWB) System ..... 3.7-1 3.7.2 Service Water (SW) System and Ultimate Heat Sink (UHS) ................... 3.7-3 3.7.3 Reactor Equipment Cooling (REC) System .................. '. ......................... 3. 7-6 3.7.4 Control Room Emergency Filter (CREF) System ................................... 3.7-8 3.7.5 Air Ejector Offgas ................................................................................... 3. 7-11 3.7.6 Spent Fuel Storage Pool Water Level ...............................................*..... 3.7-13 3.7.7 The Main'Turbine Bypass System ......................................................... 3.7-14 (continued) I Cooper ii Amendment No.~

RCS Leakage Detection Instrumentation 3.4.5 Insert 1 Required Action and -s:+ Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition not met. AN~

A,B,orC

-Gt- ~ n MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />

\ ~

--B:- All required leakage -B: Enter LCO 3.0.3. Immediately.

LJ--A

~

detection systems inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Perform a CHANNEL CHECK of required drywell In accordance with atmospheric monitoring channel. the Surveillance Frequency Control Program SR 3.4.5.2 Perform a CHANNEL FUNCTIONAL TEST of In accordance with required leakage detection instrumentation. the Surveillance Frequency Control

. Program SR 3.4.5.3 Perform a CHANNEL CALIBRATION of required

!Insert 11 C. -------------NOTE-------------- C. 1 Analyze grab samples of the Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> On ly applicable when the primary containment drywall atmospheric atmosphere.

gaseous radiation monitor is the only OPERABLE AND monitor.

C.2 Monitor RCS LEAKAGE by administrative means. Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Drywall floor drain sump flow monitoring system inoperable.

C.3 Restore drywall floor drain sump flow monitoring system to OPERABLE st,atus. 7 days

NLS2019011 Page 1 of8 Attachment 3 Proposed Technical Specifications Bases Changes (Mar){-up) -

Information Only Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 B 3.4-24 B 3.4.25 B 3.4-26 B 3 .4-26 Insert B 3.4-27 B 3.4-27 Insert B 3.4-28

RCS Leakage Detection Instrumentation B 3.4.5 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.5 RCS Leakage Detection Instrumentation BASES BACKGROUND

  • USAR Safety Design Basis (Ref. 1) requires means for detecting and, to the extent practical, identifying the location of the source of RCS LEAKAGE. Regulatory Guide 1.45 ef. 2) describes acceptable methods for selecting leakage detec 10n s [ , Revision o, Limits on LEAKAGE from the reactor coolant pressure boundary (RCPB) are required so that appropriate action can be taken before the integrity of the RCPB is impaired (Ref. 2). Leakage detection systems for the RCS are provided to alert the operators when leakage rates above normal In addition to meeting the background levels are detected and also to supply quantitative OPERABILITY measurement of leakage rate The Bases for LCO 3.4.4, "RCS requirements, the monitors Operational LEA ," 1scuss the limits on RCS LEAKAGE rates.

are typically set to provide the most sensitive ,__..,.."stems for separating the LEAKAGE of an identified source from an response without causing unidentified source are necessary to provide prompt and quantitative an excessive number of information to the operators to permit them to take immediate corrective spurious alarms. action.

LEAKAGE from the RCPB inside the drywell is detected by at least one of two independently monitored variables, such as sump flow and drywell gaseous (noble gas) and particulate radioactivity levels. The primary means of quantifying LEAKAGE in the drywell is the drywell floor drain sump flow monitoring system.

The drywell floor drain sump flow monitoring system monitors the LEAKAGE collected in the floor drain sump. This unidentified LEAKAGE consists of LEAKAGE from control rod drives, valve flanges or packings, floor drains, the Reactor Equipment Cooling System, and drywell air cooling unit condensate drains, and any LEAKAGE not collected in the drywell equipment drain sump.

)

Cooper B 3.4-24

RCS Leakage Detection Instrumentation B 3.4.5 BASES BACKGROUND (continued)

A flow transmitter in the discharge line of the drywell floor drain sump pumps provides flow indication in the control room. The pumps can also be started from the control room.

The 2-channel, drywall air monitoring system continuously monitors the primary containment atmosphere for airborne particulate and gaseous (noble gas) radioactivity. A sudden increase of radioactivity, which may be attributed to RCPB steam or reactor water LEAKAGE, is annunciated in the control room. The 2 cl'laRRel df'yVv'SHatmospl'lereparticulateaRa--

gaseous (noble gas) i:aeioactivity li!Onitoring system is not eapab!s of quaRtifyiAg LEAKAGE Fates, 13ut is SORSitive OAOU§R to iRdieate iAOFOased LEAl<AGE Fates (Ref. 3). .

APPLICABLE SAFETY ANALYSIS A threat of significant compromise to the RCPB exists if the barrier I I rate limits are set low enough to detect the LEAKAG!

single crack in the RCPB (Refs. ~AS 5). Eaeh of t R t i a R

=::

contains a crack that is large enough to propagate rapidly. LEAKAGE 3

and 4.

~ m s isdesigned 'Nith thesapability of detecting beAKAGe-less than the estaslished LEAKAGE Fate limits and providing appropriate alarm of excess LEAKA:GE iA U'le eoRtrol room.

A control room alarm allows the operators to evaluate the significance of the indicated LEAKAGE and, if necessary, shut down the reactor for further investigation and corrective action. The allowed LEAKAGE rJ3t~

are well below the rates predicted for the critical crack sizes (Ref.~ -

Therefore, these actions provide adequate response before a significant break in the RCPB can occur.

RCS leakage detection instrumentation satisfies Criterion 1 of 10 CFR 50.36(c)(2)(ii) fR-ef. 7).

  • LCO This LCO requires instruments of diverse monitoring principles to be OPERABLE to provide confidence that small amounts of unidentified LEAKAGE are detected in time to allow actions to place the plant in a safe condition, when RCS LEAKAGE indicates possible RCPB degradation.

)

Cooper B 3.4-25 11/08/18

RCS Leakage Detection Instrumentation B 3.4.5 BASES LCO (continued)

~ 4 h e loakageeetestien systems ineperab!e, monitering fer LEAKAGE in the RGP--8 is aegraaee.

APPLICABILITY In MODES 1, 2, and 3, leakage detection systems are required to be OPERABLE to support LCO 3.4.4. This Applicability is consistent with that for LCO 3.4.4.

ACTIONS With the drywell floor drain sump flow monitoring system inoperable, no other form of sampling can provide the equivalent information to quantify leakage. However, the drywall atmospheric activity monitor will provide indication of changes in leakage.

With the drywell floor drain sump flow monitoring system inoperable, but with RCS unidentified and total LEAKAGE being determined every 1;;? hours (SR 3.4.4.1 ), operation may continue for 30 days. The 30 day Completion Time of Required Action A.1 is acceptable, based on operating experience, considering the multiple forms of leakage detection that are still available.

B.1 and B.2 With both gaseous and particulate drywell atmospheric monitoring channels inoperable, grab samples of the drywall atmosphere must be taken and analyzed to provide periodic leakage information. Provided a sample is obtained and analyzed once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the plant may be operated for up to 30 days to allow restoration of at least one of the required monitors.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval provides periodic information that is adequate to detect LEAKAGE. The 30 day Completion Time for

.. _)

Cooper B 3.4-26 12J20.'18

!Insert 1!

This LCO requires two leakage detection instruments to be OPERABLE.

The drywell floor drain sump flow monitoring system is required to quantify the unidentified LEAKAGE rate from the RCS. Thus, for the system to be considered OPERABLE, the flow monitoring portion must be OPERABLE and capable of determining the leakage rate. The identification of an increase in unidentified LEAKAGE will be delayed by the time required for the unidentified LEAKAGE to travel to the drywell floor drain sump and it may take longer than one hour to detect a 1 gpm increase in unidentified LEAKAGE, depending on the origin and magnitude of the LEAKAGE. This sensitivity is acceptable for drywell sump motor OPERABILITY.

The reactor coolant contains radioactive material that, when released to the primary containment, can be detected by the gaseous or particulate drywell atmospheric radioactivity monitor. Only one of the two detectors is required to be OPERABLE. Radioactivity detection systems are included for monitoring both portable and gaseous activities because of their sensitivities and rapid responses to RCS LEAKAGE, but have rE!cognized limitations. Reactor coolant radioactivity levels will be low during initial reactor startup and for a few weeks thereafter, until activated corrosion products have been formed and fission products appear from fuel element cladding contamination and cladding defects. If there are few fuel element cladding defects and low levels of activation products, it may not be possible for the gaseous or particulate drywell radioactivity monitors to detect a 1 gpm increase within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during normal operation. However, the particulate drywell radioactivity monitor is OPERABLE when it is capable of detecting a 1 gpm increase in unidentified LEAKAGE within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> given an RCS activity equivalent to that assumed in the design calculations for the monitors. Additionally, the gaseous drywell radioactivity monitor is OPERABLE when it is capable of detecting a LEAKAGE rate of less than the LEAKAGE rate limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> given an RCS activity equivalent to that assumed in the design calculations for the monitors (Reference 6).

This LCO is satisfied when monitors of diverse measurement means are OPERABLE. Thus, the drywell floor drain sump monitoring system, in combination with a gaseous or particulate drywell atmospheric radioactivity monitor, provides an acceptable minimum.

RCS Leakage Detection Instrumentation B 3.4.5 BASES AC!IONS (continued}

restoration recognizes that at least one other form of leakage detection is

!Insert 2 available.

-6:4-ami-:-r ~, ~ 1' If any Required Action and associated Completion Time of Condition

.--------::-, ;:!;J cannot be met, the plant must be brought to a MODE in which the LCO

~ tloes not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 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 perform the actions in an orderly manner and without challenging plant systems.

With all required monitors inoperable, no required automatic means of monitoring LEAKAGE are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required.

SURVEILLANCE REQUIREMENTS

  • SR 3.4.5.1 This SR is for the performance of a CHANNEL CHECK of the required drywall atmospheric monitoring system. The check gives reasonable confidence that the channel is operating properly.* The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Cooper B 3.4-27

!Insert 21 C.1, C.2, and C.3 With the drywall floor drain sump flow monitoring system inoperable, and the drywall atmospheric particulate monitor inoperable, the only means of detecting LEAKAGE is the drywall atmospheric gaseous radiation monitor. A Note clarifies this applicability of the Condition. The drywall atmospheric gaseous radiation monitor typically cannot detect a 1 gpm leak within one hour when RCS activity is low. In addition this configuration does not provide the required diverse means of leakage detection. Indirect methods of monitoring RCS leakage must be implemented. Grab samples of the primary containment atmosphere must be taken and analyzed and monitoring of RCS leakage by administrative means must be performed every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to provide alternate periodic information.

Administrative means of monitoring RCS leakage include monitoring and trending parameters that may indicate an increase in RCS leakage. There are diverse alternative mechanisms from which appropriate indicators may be selected based on plant conditions. It is not necessary to utili~e all of these methods, but a method or methods should be selected considering the current plant conditions and historical or expected sources of unidentified leakage. Some administrative methods available are drywall equipment sump temperature, suppression pool water level, primary containment pressure, and primary containment temperature. These indications, coupled with the atmospheric grab samples, are sufficient to alert the operating staff to an unexpected increase in unidentified LEAKAGE.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval is sufficient to detect increasing RCS leakage. The Required Action provides 7 days to restore another RCS leakage monitor to OPERABLE status to regain the intended leakage detection diversity. The 7 day Completion Time ensures that the plant will not be operated in a degraded configuration for a lengthy time period.

RCS Leakage Detection Instrumentation B 3.4.5 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.5.2 This SR is for the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation. The test ensures that the monitors can perform their function in the desired manner. The test also verifies the alarm setpoint and relative accuracy of the instrument string. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.5.3 This SR is for the performance of a CHANNEL CALIBRATION of required leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string. The Surveillance Frequency is.

controlled under the Surveillance Frequency Control Program.

Revision 0, "Reactor Coolant Pressure REFERENCES 1. USAR, Section IV-10.2.

Boundary Leakage Detection Systems,"

2. Regulatory Guide 1.45, ay 1973.
3. USAR, Seetiofl l'.Jw10.3.

GEAP-5620, "Failure Behavior in ASTM A 106B Pipes Containing Axial Through-Wall Flaws," April 1968.

NUREG-75/067, "Investigation and Evaluation of Cracking in Austetic Stainless Steel Piping of Boiling Water Reactors,"

October 1975.

USAR, Section IV-10.3.2.

7. 1g C~R eG.3e(c)~(.U}.-

\'.S: !6. USAR, Section IV-10.3. !

Cooper B 3.4-28 0§/17.'17

NLS2019011 Page 1 of 16 Attachment 4 Revised Technical Specifications Pages Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 TOC page ii 3.4-11 3.4-12 3.4-13*

3.4-14*

3.4-15*

3.4-16*

3.4-17*

3.4-18*

3.4-19*

3.4-20*

3.4-21

  • 3.4-22*

3.4-23*

3.4-24*

  • included due to repagination only

TABLE OF CONTENTS (continued) 3.4 REACTOR COOLANT SYSTEM (RCS) ..... ..... ... .... .. .... .. ... .. ...... ........ .... .. ..... 3.4-1 3.4.1 Recirculation Loops Operating ......... ...... .. .... .. .... .. ... ... .. ... .. .... ... ... .... .... .. . 3.4-1 3.4.2 Jet Pumps ..... ...... .... ... .... .......... ... .... ... ..... ......... .. ..... .. ...... .. ... .... ... ..... .... .. 3.4-4 3.4.3 Safety/Relief Valves (SRVs) and Safety Valves (SVs) .... .... .. ... .. ... .. .. ..... 3.4-6 3.4.4 RCS Operational LEAKAGE .. ... .... ........ ... ........ ... .. ..... ... ........ .. .. ....... ... ... 3.4-8 3.4.5 RCS Leakage Detection Instrumentation .... ... ......... .... .... ...... .. .. .......... ... 3.4-10 3.4.6 RCS Specific Activity ....... .. ..... ............ .... ... .. ..... ... ....... ....... ..... ... ...... ...... 3.4-13 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown ....... .. ... .... .. ....... .. .. ... ... ............... ............ .. .. .. 3.4-15 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown ... .. .... ....... .. .... ... .. ... .... ........ .... ..... .... ...... ... .. 3.4-18 3.4.9 RCS Pressure and Temperature (PIT) Limits .... .. .. .. ... ... .. .. .. .... ........ ... .... 3.4-20 3.4.10 Reactor Steam Dome Pressure ..... ........ .. .... ... ....... ... ... ..... .. .. .. ........... ... . 3.4-24 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ..... ... .. ...... .. ..... ............ .... ... ..... ...... .... ......... .. .... 3.5-1 3.5.1 ECCS - Operating ....... ..... .. ........................ ................ ......... ....... ... ........ . 3.5-1 3.5.2 RPV Water Inventory Control .... ...... ....... ..... .. .. .. ... .... ...... ..... ... ........... .. ... 3.5-7 3.5.3 RCIC System ............ .......... ... ...... ....... .... ...... ...... ........ ........... .... ... ... .. .... 3.5-12 3.6 CONTAINMENT SYSTEMS ... .. .... ... ........ .... .. .... .......... ..... .. ... ... ....... ...... ...... 3.6-1 3.6 .1.1 Primary Containment ... ... ... .... .... ..... .. ... ... .. .... ... ... .. .... .. .... ...... ... ... .. .. ....... 3.6-1 3.6 .1.2 Primary Containment Air Lock ... ... .... ....... ............ .... ..... ...... ...... .... ..... ..... 3.6-3 3.6.1 .3 Primary Containment Isolation Valves (PCIVs) ...... ... .. .. ... ... ..... ... ..... ..... . 3.6-8 3.6.1.4 Drywell Pressure ... .. .... .... ... .. ..... ..... ..... .... .. ... .................. .............. ..... ..... 3.6-16 3.6.1.5 Drywell Air Temperature .. .. ........ ............... ............ ....... ............ ..... ..... .... 3.6-17 3.6.1.6 Low-Low Set (LLS) Valves ... ....... .. ..... ..... ... .... .... .. .......... .. ..... ..... ... .. ..... .. 3.6-18 3.6.1 .7 Reactor Building-to-Suppression Chamber Vacuum Breakers ..... ... ....... 3.6-20 3.6.1.8 Suppression-Chamber-to-Drywell Vacuum Breakers .. .. .. .. .... ..... ... ... ..... . 3.6-23 3.6 .1.9 Residual Heat Removal (RHR) Containment Spray ... .. .... ... .. ... .... .. .... ... . 3.6-25 3.6.2.1 Suppression Pool Average Temperature .... ..... ..... .... ....... .......... ... ......... 3.6-27 3.6.2.2 Suppression Pool Water Level .... ......... .. .. ... ...... ....... ........ .... ... .... ...... .... . 3.6-30 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling .... ... .. .. ... ....... 3.6-31 3.6.3.1 Primary Containment Oxygen Concentration ...... .... ... .... ..... .. ... ... ... ... ... .. 3.6-33 3.6.4.1 Secondary Containment.. ....... .. ......... ... .. ..... ...... .. .. ...... .. ......... .... .. .......... 3.6-34 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) ........ ....... .... ... ... ........ . 3.6-36 3.6.4.3 Standby Gas Treatment (SGT) System .. ....... ...... .. .. ... ......... ............. ...... 3.6-40 3.7 PLANT SYSTEMS .. ... ... ... ........ .. .... .. ..... .. ...... ....... ........ .... ........... ... ... .... .... .. . 3.7-1 3.7.1 Residual Heat Removal Service Water Booster (RHRSWB) System .... . 3.7-1 3.7.2 Service Water (SW) System and Ultimate Heat Sink (UHS) .. ....... .... ..... . 3.7-3 3.7.3 Reactor Equipment Cooling (REC) System .. ...... .. ..... ......... ......... .... ....... 3. 7-6 3.7.4 Control Room Emergency Filter (CREF) System .... ..... .. ... ........ .. ... .. ... ... 3.7-8 3.7.5 Air Ejector Offgas ..... ...... ... ..... .. ..... .................. ..... ...... ... ..... .. .... ... ... ...... .. 3. 7-11 3.7.6 Spent Fuel Storage Pool Water Level .... ... ..... ... .. ....... .. ..... .... .... ....... ..... . 3.7-13 3.7.7 The Main Turbine Bypass System .. ..... .......... .......... ... .. ......... .... .. ..... .... . 3. 7-14 (continued)

Cooper ii Amendment No.

RCS Leakage Detection Instrumentation 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. --------------NOTE--------------- C.1 Analyze grab samples of Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Only applicable when the the primary containment drywell atmospheric atmosphere.

gaseous radiation monitor is the only OPERABLE AND monitor.

C.2 Monitor RCS LEAKAGE Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by administrative means.

Drywell floor drain sump flow monitoring system AND inoperable.

C.3 Restore drywell floor drain 7 days sump flow monitoring system to OPERABLE status.

D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, B, or C not AND met.

D.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> E. All required leakage E.1 Enter LCO 3.0.3. Immediately detection systems inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Perform a CHANNEL CHECK of required drywell In accordance with atmospheric monitoring channel. the Surveillance Frequency Control Program (continued) I Cooper 3.4-11 Amendment No.

RCS Leakage Detection Instrumentation 3.4.5 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.5.2 Perform a CHANNEL FUNCTIONAL TEST of In accordance with required leakage detection instrumentation . the Surveillance Frequency Control Program SR 3.4.5.3 Perform a CHANNEL CALIBRATION of required In accordance with leakage detection instrumentation. the Surveillance Frequency Control Program Cooper 3.4-12 Amendment No.

RCS Specific Activity 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Specific Activity LCO 3.4.6 The specific activity of the reactor coolant shall be limited to DOSE EQUIVALENT 1-131 specific activity::;; 0.2 µCi/gm.

APPLICABILITY: MODE 1, MODES 2 and 3 with any main steam line not isolated.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor coolant specific -------------------- .NOTE-------------------

activity> 0.2 µCi/gm and :5: LCO 3.0.4.c is applicable.

4.0 µCi/gm DOSE ------------------------------------------------

EQUIVALENT 1-131.

A.1 Determine DOSE Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> EQUIVALENT 1-131.

AND A.2 Restore DOSE 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> EQUIVALENT 1-131 to within limits.

B. Required Action and B.1 Determine DOSE* Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion Time EQUIVALENT 1-131.

of Condition A not met.

AND OR B.2.1 Isolate all main steam 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Reactor Coolant specific lines.

activity > 4.0 µCi/gm DOSE EQUIVALENT 1-131. OR (continued)

Cooper 3.4-13 Amendment No.

RCS Specific Activity 3.4.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2.2.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B.2.2.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 ----------.--------------------NOTE-------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 In accordance with specific activity is ~ 0.2 µCi/gm. the Surveillance Frequency Control Program Cooper 3.4-14 Amendment No.

RHR Shutdown Cooling System - Hot Shutdown 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4. 7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown LCO 3.4.7 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

NOTES-------------------------------------------

1. Both RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.
2. One RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of Surveillances.

APPLICABILITY: MODE 3, with reactor steam dome pressure less than the shutdown cooling permissive pressure.

ACTIONS

. ---------NOTE------------------------------------------------------------

Separate Condition entry is allowed for each RHR shutdown cooling subsystem.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or two RHR shutdown A.1 Initiate action to restore Immediately cooling subsystems RHR shutdown cooling inoperable. subsystem( s) to OPERABLE status.

AND (continued)

Cooper 3.4-15 Amendment No.

RHR Shutdown Cooling System - Hot Shutdown

. 3.4.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 Verify an alternate method 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of decay heat removal is available for each inoperable RHR shutdown cooling subsystem.

AND A.3 Be in MODE 4. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> B. No RHR shutdown cooling 8.1 lnhiate action to restore Immediately subsystem in operation. one RHR shutdown cooling subsystem or one AND recirculation pump to operation.

No recirculation pump in operation. AND 8.2 Verify reactor coolant 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from circulation by an alternate discovery of no method. reactor coolant circulation AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND 8.3 Monitor reactor coolant Once per hour temperature and pressure.

Cooper 3.4-16 Amendment No.

RHR Shutdown Cooling System - Hot Shutdown 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 -------------------------------N()TE---~---------------------------

Not required to be met until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after reactor steam dome pressure is less than the shutdown cooling permissive pressure.

Verify one RHR shutdown cooling subsystem or In accordance with recirculation pump is operating. the Surveillance Frequency Control Program Cooper 3.4-17 Amendment No.

RHR Shutdown Cooling System - Cold Shutdown 3.4.8 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown LCO 3.4.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

NOTES------------------------------------------

1. Both RHR shutdown cooling subsystems and recirculation pumps may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.
2. One RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the performance of Surveillances.

APPLICABILITY: MODE 4.

ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each shutdown cooling subsystem.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or two RHR shutdown A.1 Verify an alternate method 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> cooling subsystems of decay heat removal is inoperable. available for each inoperable RHR shutdown cooling subsystem. Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter (continued)

Cooper 3.4-18 Amendment No.

RHR Shutdown Cooling System - Cold Shutdown 3.4.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. No RHR shutdown cooling B.1 Verify reactor coolant 1 hourfrom subsystem in operation. circulating by an alternate discovery of no method. reactor AND coolant circulation No recirculation pump in AND operation.

Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND B.2 Monitor reactor coolant Once per hour temperature.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one RHR shutdown cooling subsystem or In accordance with recirculation pump is operating. the Surveillance Frequency Control Program

, Cooper 3.4-19 Amendment No.

RCS PIT Limits 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 RCS Pressure and Temperature (PIT) Limits LCO 3.4.9 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within the limits specified in the PTLR.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ---------------NOTE-------------- A.1 Restore parameter( s) to 30 minutes Required Action A.2 shall be within limits.

completed if this Condition is entered. AND A.2 Determine RCS is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Requirements of the LCO acceptable for continued not met in MODE 1, 2, or 3. operation.

B. Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not met. AND 8.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

Cooper 3.4-20 Amendment No.

RCS PIT Limits 3.4.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ---------------NOTE-------------- C.1 Initiate action to restore Immediately Required Action C.2 shall parameter(s) to within be completed if this limits.

Condition is entered.

AND Requirements of the LCO C.2 Determine RCS is Prior to entering not met in other than acceptable for operation. MODE 2 or 3.

MODES 1, 2, and 3.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 --~---------------------------NOTE--------------------------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify: In accordance with the Surveillance

a. RCS pressure and RCS temperature are Frequency Control within the applicable limits specified in the Program curves in the PTLR; and
b. RCS heatup and cooldown rates are within limits specified in the PTLR.

(continued)

Cooper 3.4-21 Amendment No.

RCS P/T Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.2 Verify RCS pressure and RCS temperature are within Once within the criticality limits specified in the PTLR. 15 minutes prior to control rod withdrawal for the purpose of achieving criticality SR 3.4.9.3 --------------------------------1\lOTE-------------------------------

Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.

Verify the difference between the bottom head Once within coolant temperature and the reactor pressure vessel 15 minutes prior to (RPV) coolant temperature is within the limits each startup of a specified in the PTLR.

  • recirculation pump SR 3.4.9.4 -------------------------------~NOTE------------------------------

Only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup.

Verify the difference between the reactor coolant Once within temperature in the recirculation loop to be started and 15 minutes prior to the RPV coolant temperature is within the limits each startup of a specified in the PTLR. recirculation pump (continued)

Cooper 3.4-22 Amendment No.

RCS PIT Limits 3.4.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.9.5 -------------------------------NC>TE--------------------------------

()nly required to be performed when tensioning the reactor vessel head bolting studs.

Verify reactor vessel flange and head flange In accordance with temperatures are within the limits specified in the the Surveillance PTLR. Frequency Control Program SR 3.4.9.6 -------------------------------NC>TE--------------------------------

Not required to be performed until 30 minutes after RCS temperature s 80°F in M()DE 4.

Verify reactor vessel flange and head flange In accordance with temperatures are within the limits specified in the the Surveillance PTLR. Frequency Control Program SR 3.4.9.7 -------------------------------NC>TE-* ------------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature :s; 90°F in MC>DE 4.

, Verify reactor vessel flange and head flange In accordance with temperatures are. within the limits specified in the the Surveillance PTLR. Frequency Control Program Cooper 3.4-23 Amendment No.

Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure LCO 3.4.10 The reactor steam dome pressure shall be s 1020 psig.

APPLICABILITY: MODES 1 and 2.

ACTIONS CON.DITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within limit. dome pressure to within limit.

B. Required Action and 8.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify reactor stean:i dome pressure is s 1020 psig. lri accordance with the Surveillance Frequency Control Program Cooper 3.4-24 Amendment No.

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