Text

Response to Supplemental Information Needed for Acceptance of Requested Licensing Action Re-Renewed Facility Operating License and Technical Specification Clean-Up
	ML18219C797
Person / Time
Site:	Columbia
Issue date:	08/07/2018
From:	Schuetz R; Energy Northwest
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	EPID L-2018-LLA-0178, GO2-18-102
	Download: ML18219C797 (114)
	v • d • e

ENERGY Robert E. Schuetz Vice President, Operations P.O. Box 968, Mail Drop PE23 NORTHWEST Richland, WA 99352-0968 Ph. 509-377-2425 F. 509-377-4674 reschuetz@energy-northwest.com August 7, 2018 GO2-18-102 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397 RESPONSE TO SUPPLEMENTAL INFORMATION NEEDED FOR ACCEPTANCE OF REQUESTED LICENSING ACTION RE: RENEWED FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATION CLEAN-UP

References:

1. Letter from R.E. Schuetz, Energy Northwest to Nuclear Regulatory Commission (NRC), Columbia Generating Station, Docket No. 50-397 License Amendment Request to Clean-up Operating License and Appendix A Technical Specifications, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18163A351), dated June 12, 2018.

2. Letter from L.J. Klos, NRC to B.J. Sawatzke, Energy Northwest, Columbia Generating Station - Supplemental Information Needed for Acceptance of Requested Licensing Action Re: Renewed Facility Operating License and Technical Specification Clean-up (EPID L-2018-LLA-0178), ADAMS Accession No. ML18206A324, dated July 31, 2018.

Dear Sir or Madam:

By Reference 1 Energy Northwest submitted a license amendment request (LAR) for Columbia Generating Station (Columbia). The amendment proposes a number of clean-up changes to the OL and TS, including editorial changes and the removal of obsolete TS and OL information. By Reference 2 the NRC requested supplemental information be provided to enable the NRC staff to make an independent assessment regarding the acceptability of the proposed amendment. Enclosures 1 through 5 of of this letter replace and supersede in their entirety Enclosures 1 through 5 of the original submittal in Reference 1 dated June 12, 2018.

GO2-18-102 Page 2 of 2 No new commitments are being made by this letter or the enclosures. Additionally, the No Significant Hazards Consideration determination in the original submittal is not altered by the information provided in this response. Approval of the proposed amendment is still requested within one year of the date of the submittal. Once approved, Energy Northwest shall implement the amendment within 90 days.

If there are any questions or if additional information is needed, please contact Mr. R.M.

Garcia, Licensing Supervisor, at 509-377-8463.

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

Executed this 7 day of Au 9:)"'~-\- , 2018.

cc: NRC RIV Regional Administrator NRC NRR Project Manager NRC Senior Resident lnspector/988C CD Sonoda - BPA/1399 (email)

EFSECutc.wa.gov - EFSEC (email)

RR Cowley-WDOH (email)

WA Horin - Winston & Strawn

GO2-18-102 SUPPLEMENTAL INFORMATION NEEDED LICENSE AMENDMENT REQUEST REGARDING RENEWED FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATION CLEAN-UP ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397 By letter dated June 12, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18163A351), Energy Northwest (the licensee) submitted a license amendment request (LAR) for Columbia Generating Station (Columbia). The proposed amendment would remove the Table of Contents from the Technical Specifications (TSs) and place it under licensee control, and clean-up the renewed facility operating license (RFOL) and TSs, including editorial changes, the removal of obsolete TS information, renumbering TS pages and the removal of obsolete license conditions and attachments. The NRC staff has concluded that supplemental information is required in order to make the application complete, as discussed below.

Background

The LAR's cover letter introduction states, in part, that this request is an administrative change that does not "result in changes to the technical or operating requirements" of Columbia. The LAR further states, in Section 2.2.4, that as low as reasonable achievable (ALARA) reviews were made of areas where shield walls were deferred (not constructed per Columbia Amendment No. 7) and these results were considered acceptable.

Reference 12 of the LAR (Inspection Report 05000397 /2016003 for Columbia; ADAMS Accession No. ML16302A315), states that in 2015 certain shield wall location areas found in Attachment 3, "List of Shield Walls," of RFOL NPF-21 for Columbia (where a shield wall does not physically exist) have exceeded radiation and ALARA controls.

These areas are locations that are also related to the RFOL License Condition 2.C.(11),

"Shield Wall Deferral (Section 12.3.2, SSER #4, License Amendment #7)."

Reference 12 of the LAR, states that non-compliance in Attachment 3 areas related to the license condition stated above, have occurred in 2010 and 2015, and that in 2015, a licensee-identified non-cited violation for this non-compliance issue was issued, which subsequently created two licensee corrective action program entries (Action requests (ARs) 00354266 and 0035320).

Reference 12 of the LAR, also states that the intent of the ARs included an "update [to Columbia's] [license] condition to accurately reflect current operations." Therefore, the amendment's proposed changes are to remove License Condition 2.C(11) and from the TSs. Reference 12 of the LAR, further states that an ALARA program was not practiced in full for areas of Attachment 3 related to the license condition from 2009 to 2015.

The amendment's proposed changes to remove License Condition 2.C(11), concerning Shield Wall deferment (Columbia's Amendment No. 7) and the associated areas stated in Attachment 3, would then omit documentation of Columbia's current licensing basis.

GO2-18-102 Specifically, the proposed changes would therefore not reflect Columbia's current operations nor the technical and operating requirements associated with Columbia's Amendment No. 7 and the areas related to shield wall deferment.

Supplemental Information Requested:

1. The licensee is requested to provide a supplement that either; a) Retains and documents licensing basis continuity for the items above rather than deleting them per the current submittal; or b) Provides a full justification for the intended change that includes the complete licensing history, documentation and discussion that would support the proposed change while still representing the full technical and operating arrangement of Columbia. Additionally, a full justification should fully document, as proposed in the amendment, that License Condition 2.C(11) is obsolete.

Response to Supplemental Information Request:

Energy Northwest submits Enclosures 1 through 5 of this Attachment to replace and supersede in their entirety Enclosures 1 through 5 of the original submittal in Reference 1 (ADAMS Accession No. ML18163A351) dated June 12, 2018. The superseding Enclosures fully document justification for the requested LAR as amended by this supplement.

GO2-18-057 Page 1 of 10 Evaluation of Proposed Operating License and Technical Specification Changes 1.0

SUMMARY

DESCRIPTION This evaluation supports a License Amendment Request (LAR) to the Columbia Generating Station (Columbia) Operating License (OL) NPF-21 and Appendix A, Technical Specifications (TS). The proposed changes are clean-ups of the OL and TS.

They are administrative and editorial and will not result in any change to operating requirements. This amendment is requested to remove the Table of Contents (TOC) from the TS and place it under licensee control. Additionally, this amendment proposes a number of editorial changes to the OL and TS, including, but not limited to, removal of an obsolete OL condition and OL attachments, the removal of obsolete TS information, and renumbering TS pages. Neither the proposed administrative changes nor the proposed editorial changes result in changes to technical or operating requirements.

The specific sections of the TS affected by the changes are listed below:

Page i through iv TOC Section 1.1 Definitions Section 3.2.4 APRM Gain and Setpoint Section 3.3.1.1 RPS Instrumentation Section 3.3.1.3 OPRM Instrumentation Section 3.3.2.1 Control Rod Block Instrumentation Section 3.4.1 Recirculation Loops Operating Section 3.10.8 SDM Test - Refueling Section 5.6.3 Reporting Requirements Various Sections 3.1.6, 3.3.6.1, 3.3.8.1, 3.5.1, and 5.5.4 contain typographical errors that will be corrected in this amendment.

The specific condition and attachments of the OL affected by the changes are:

Condition 2.C.(33) Control Room Envelope Habitability Program (CRE) Deleted Deleted Implementation of this LAR will result in no physical modification to the plant. This proposed change has no adverse effect on the plant or plant safety.

GO2-18-057 Page 2 of 10 2.0 DETAILED DESCRIPTION 2.1 Current Technical Specification and Operating License Energy Northwest proposes a clean-up to Columbias OL and TS to remove obsolete information and correct administrative errors. The changes may cause repagination of certain sections of both documents. Those changes are incorporated into this LAR.

Energy Northwest proposes to correct typographical errors in various sections of TS.

Also the TOC will be removed from the TS and placed under licensee control. Energy Northwest is addressing the reasoning and descriptions for the proposed changes below.

2.2. Reason and Description for the Proposed Changes 2.2.1 Removal of TOC from TS The proposed change would revise the OL to remove the TOC from the TS. The TOC for the TS is not being eliminated, rather, following approval of this LAR, responsibility for maintenance and issuance of updates to the TS TOC will transfer from the Nuclear Regulatory Commission (NRC) to Energy Northwest. The TOC will no longer be included in the NRC issued TS and as such will no longer be part of the TS (Appendix A to the Operating License). The TOC for the TS will be maintained under Energy Northwests control. The TOC will be issued by Energy Northwest in conjunction with the implementation of future NRC approved TS amendments.

Placing the TOC under licensee control eliminates the regulatory burden of submitting TOC pages for NRC review and allows timely administrative corrections and improvements to the TOC without NRC review and approval.

2.2.2 Amendment 226, Implementation of Power Range Neutron Monitoring/Average Power Range Monitor Rod Block Monitor Technical Specifications/Maximum Extended Load Line Limit Analysis (PRNM/ARTS/MELLLA) Clean-up On January 31, 2014, Energy Northwest received license Amendment 226, allowing the implementation of the Power Range Neutron Monitoring (PRNM) upgrade, ML13317B623. (Reference 1) The LAR contained changes to several TS. The license amendment was effective on January 31, 2014, however it would not be implemented until prior to plant startup following refueling outage 22. The refueling outage was scheduled to begin in May 2015.

In support of the license amendment request, Energy Northwest submitted a Response to Request for Additional Information Regarding License Amendment Request to Implement PRNM/ARTS/MELLLA. (Reference 2). Enclosure 4 of that submittal was the Revision to Description of Proposed Technical Specification Changes. The changes made to the proposed TS were required to accommodate the implementation schedule. TS 3.3.1.1, 3.3.1.3, 3.3.2.1, 3.2.4, 3.4.1, and 3.10.8 were modified to reflect

GO2-18-057 Page 3 of 10 the applicability of prior to implementation of PRNM upgrade or after implementation of PRNM upgrade, as appropriate. The header of each specification included the same modification.

In that response, Energy Northwest stated that once the PRNM upgrade was installed in the plant, a TS change would be submitted to delete the obsolete Specifications and portions thereof. Amendment 226 (Reference 1) contains recognition that the information below would be removed from Columbias TS after PRNM/ARTS/MELLLA implementation.

1. Specifications 3.2.4 Power Distribution Limits and 3.3.1.3 OPRM Instrumentation will be deleted. These Specifications are obsolete with the implementation of the PRNM upgrade.

2. The Definition of Maximum Fraction of Limiting Power Density (MFLPD) will be deleted. Since TS 3.2.4 is being deleted, this Definition is no longer in use at Columbia.

3. Specifications 3.3.1.1, 3.3.2.1, 3.4.1, and 3.10.8 contain pages that are applicable prior to implementation . . . and after implementation . . ..

Each portion of these Specifications containing the header Prior to implementation of PRNM Upgrade will be deleted. TS pages 3.3.1.1-1 through 8, 3.3.2.1-1 through 6, 3.4.1-1 through 2 and 3.10.8-1 through 4 will be deleted.

Each of the remaining Specifications containing the header language After implementation of PRNM Upgrade will have this language deleted from the header. The language after implementation of Power Range Neutron Monitor (PRNM) upgrade will be deleted from the APPLICABILITY. The remaining TS pages will be renumbered to 3.3.1.1-1 through 10, 3.3.2.1-1 through 7, 3.4.1-1 through 2 and 3.10.8-1 through 3.

4. Bullet a.4 of TS 5.6.3 will be deleted. The language after the PRNM upgrade will be deleted from Bullets a.5 and a.6 of TS 5.6.3.

2.2.3 Correction of Typographical Errors The correction of typographical errors in the TS are detailed in Section 3.3 below. The corrections are supported by the guidance contained in the Writer's Guide for Plant-Specific Improved Technical Specifications (Reference 3).

GO2-18-057 Page 4 of 10 2.2.4 Removal of License Condition 2.C. (33) - Control Room Envelope (CRE)

Habitability Program The first performances of the associated CRE habitability surveillance, assessment and measurement are complete, see Table 3.5 below. Successive performances of the surveillances, assessments, and measurements will follow the respective frequencies identified in TS 5.5.14.c and 5.5.14.d. With the first performances complete for Columbia, the OL Conditions are no longer required and can be removed.

2.2.5 Removal of Attachment 1 and Attachment 2 Columbias OL was modified to delete the contents of Attachments 1 and 2 with Amendment 223, on March 30, 2012, but the attachments, themselves, were retained (Reference 4). The proposed change will delete these attachments.

3.0 TECHNICAL EVALUATION

The proposed changes to Columbias OL and TS are either administrative or editorial and do not affect how plant equipment is operated or maintained. No changes to the physical plant or analytical methods are described and there are no impacts on the updated Final Safety Analysis Report (FSAR) accident analysis.

3.1 Removal of TOC from TS The TOC does not meet the criteria specified in 10 CFR 50.36 requiring its inclusion within a plant's TS. 10 CFR 50.36(b) states:

"Each license authorizing operation of a production or utilization facility of a type described in § 50.21 or § 50.22 will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to § 50.34."

Title 10 CFR 50.36(c) states the "Technical specifications will include items in the following categories." Review of 10 CFR 50.36 indicates that a TOC was not listed in the regulation as one of the categories.

Title 10 CFR 50.36(a) indicates that a license application may provide other information associated with the TS, and gives an example of this information, i.e., the TS Bases, but 10 CFR 50.36(a) also clearly indicates that the Bases are not a part of the TS.

"Each applicant for a license ... shall include in his application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications."

GO2-18-057 Page 5 of 10 The TOC references the page number of each Specification in the TS and does not contain any technical information required by 10 CFR 50.36. Since the TOC does not include information required to be in the TS by 10 CFR 50.36, inclusion of a TOC within the TS is optional. Removal of the TOC from the TS is an administrative change and is acceptable.

Additionally, the "Writer's Guide for Plant-Specific Improved Technical Specifications

(Reference 3) was reviewed for guidance. The writer's guide refers to the TOC as "Technical Specification Front Matter," which also includes the Title Page and List of Effective Pages. These portions of the TS are under licensee control, as is proposed for the TOC.

The TS TOC will be maintained, revised, and distributed in accordance with Columbias administrative procedures. Holders of copies of the TS, including the NRC, will continue to receive periodic updates of the TOC pages.

3.2 Amendment 226, Implementation of Power Range Neutron Monitoring/Average Power Range Monitor Rod Block Monitor Technical Specifications/Maximum Extended Load Line Limit Analysis (PRNM/ARTS/MELLLA) Clean-up In May and June 2015, Columbia conducted refueling outage 22. During that outage, the PRNM upgrade was implemented. Since the PRNM upgrade has been implemented, the TS pages that refer to prior to implementation . . . are obsolete and can be deleted.

Additionally, two typographical errors were identified in TS 3.3.1.1 and are being corrected in the LAR. These corrections are contained in the TS markups related to the Implementation of PRNM/ARTS/MELLLA.

1. TS 3.3.1.1, ACTION I.2 - the period is missing after the word OPERABLE.

2. TS Table 3.3.1.1-1 (page 2 of 4) - there are extra hard returns after the SURVEILLANCE REQUIREMENTS for c. Neutron Flux - High and d. Inop.

3.3 Correction of Typographical Errors The correction of typographical errors in the TS are detailed below.

1. TS 3.1.6, ACTION B - the period is missing after B.

2. TS Surveillance Requirement (SR) 3.3.6.1.5 - the period is missing after the word CALIBRATION.

3. TS Table 3.3.6.1 - the acronyms RWCU, SLC and RHR SDC should be spelled out in first use within the Table: Reactor Water Cleanup (RWCU), Standby Liquid Control (SLC), and Residual Heat Removal (RHR) Shutdown Cooling (SDC).

GO2-18-057 Page 6 of 10

4. TS SR 3.3.8.1.3 - the period is missing after the word CALIBRATION.

5. TS 3.5.1 NOTE - the acronym HPCS should be spelled out in first use within the TS: High Pressure Core Spray (HPCS).

6. TS 5.5.4.j and k - the and should be deleted from TS 5.5.4.j. The period should be deleted in TS 5.5.4.k and replaced with and ;.

3.4 Removal of License Condition 2.C. (33) - Control Room Envelope (CRE)

Habitability Program Table 3.5 CRE Habitability Completion License Condition Specification Completion Date 2.C.(33)(a) 5.5.14.c(i) 11/17/2010 2.C.(33)(b) 5.5.14.c(ii) 06/24/2010 2.C.(33)(c) 5.5.14.d - System A 03/18/2010 2.C.(33)(c) 5.5.14.d - System B 12/06/2011 3.5 Removal of Attachment 1 and Attachment 2 Removal of Attachments 1 and 2 is administrative. The License Conditions related to these Attachments have previously been deleted.

3.6 Impact on Submittals under Review by the NRC Energy Northwest has submitted a LAR to adopt Technical Specification Task Force (TSTF) TSTF-542, Reactor Pressure Vessel Water Inventory Control. TSTF-542 includes a change to TS 3.3.6.1 Primary Containment Isolation Instrumentation Table 3.3.6.1-1. The change deletes information in Function 5.d and footnote d. The present LAR spells out Residual Heat Removal (RHR) Shutdown Cooling (SDC) in the first instance of use within Table 3.3.6.1-1, which is Function 5. The changes are unrelated nor does one impact the other. Thus, the two amendment requests are not linked.

Energy Northwest has submitted a LAR to adopt TSTF-551, "Revise Secondary Containment Surveillance Requirements. This change does not affect any of the same TS pages discussed in the present LAR. Therefore, the two amendment requests are not linked.

4.0 REGULATORY EVLAUATION The Columbia FSAR Chapter 3 provides detailed discussion of Columbias compliance with the applicable regulatory requirements and guidance.

The proposed OL and TS amendment is administrative in nature and:

GO2-18-057 Page 7 of 10 Does not result in any change in the qualifications of any component; and Does not result in the reclassification of any components status in the areas of shared, safety-related, independency, redundancy, and physical or electrical separation.

4.1 Applicable Regulatory Requirements and Guidance The proposed changes to the Columbia OL and TS are either administrative or editorial and do not affect any regulatory requirements or guidance. These changes do not affect how plant equipment is operated or maintained and there are no changes to the physical plant or analytical methods. Therefore, there are no impacts on the FSAR accident analysis.

5.0 Precedent A license amendment was issued to the Monticello Nuclear Generating Plant to remove the TOC from the TS (Reference 5) on November 8, 2007. A license amendment was issued for the Waterford Steam Electric Station to remove the TS Index (which corresponds to the TOC) from the TS on May 9, 2005 (Reference 6). Also, as discussed in the response to an NRC request for additional information regarding Waterfords LAR (Reference 7), the Arkansas Nuclear One (ANO) and Grand Gulf TS issued following conversion to the Improved Standard Technical Specifications did not include a TOC. Both ANO and Grand Gulf maintain the TOC under their administrative procedures and periodically provide updated pages to controlled copy holders.

A license amendment was issued to Byron and Braidwood Stations on July 5, 2017, to remove the license conditions related to the License Amendments for TSTF-448 (Control Room Envelope (CRE) Habitability Program) after implementation. (Reference 8). Although these plants are Pressurized Water Reactors (PWRs) while Columbia is a Boiling Water Reactor (BWR), the OL implementing language of TSTF-448 for the CRE Habitability Program was the same for the reactor types.

A license amendment was issued to Columbia Generating Station on March 30, 2012, that allowed clean-up of the OL by removing obsolete license conditions. (Reference 4).

6.0 No Significant Hazards Consideration Determination Energy Northwest has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No.

GO2-18-057 Page 8 of 10 The impacts of these administrative changes do not affect how plant equipment is operated or maintained. The proposed changes do not impact the intent or substance of the Operating License (OL) or Technical Specifications (TS). There are no changes to the physical plant or analytical methods.

The proposed amendment involves administrative and editorial changes only. The proposed amendment does not impact any accident initiators, analyzed events, or assumed mitigation of accident or transient events. The proposed changes do not involve the addition or removal of any equipment or any design changes to the facility. The proposed changes do not affect any plant operations, design functions, or analyses that verify the capability of structures, systems, and components (SSCs) to perform a design function. The proposed changes do not change any of the accidents previously evaluated in the updated Final Safety Analysis Report (FSAR).

The proposed changes do not affect SSCs, operating procedures, and administrative controls that have the function of preventing or mitigating any of these accidents.

Therefore, the proposed changes do not represent 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 amendment only involves administrative and editorial changes. No actual plant equipment or accident analyses will be affected by the proposed changes. The proposed changes will not change the design function or operation of any SSCs. The proposed changes will not result in any new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.

The proposed amendment does not impact any accident initiators, analyzed events, or assumed mitigation of accident or transient events.

Therefore, this proposed changes do not create the possibility of an accident of a new or different kind than previously evaluated.

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

Response: No The proposed amendment only involves administrative and editorial changes. The proposed changes do not involve any physical changes to the plant or alter the manner in which plant systems are operated, maintained, modified, tested, or inspected. The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined.

The safety analysis acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in a configuration outside the

GO2-18-057 Page 9 of 10 design basis. The proposed changes do not adversely affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, Energy Northwest concludes that the proposed amendment to the Columbia OL and TS does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

7.0 CONCLUSION

S 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 applicable regulations as identified herein, 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.

8.0 ENVIRONMENTAL CONSIDERATION

Energy Northwest has determined that the proposed amendment would not change requirements with respect to installation or use of a facility component located within Columbia's restricted area, as defined in 10 CFR 20, nor would it change an inspection or surveillance requirement. Energy Northwest has evaluated the proposed changes and has determined that the changes do 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 occupational radiation exposure. Accordingly, the proposed changes meets the eligibility criteria for categorical exclusion in accordance with 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 amendment.

9.0 REFERENCES

1. Letter from C.F. Lyon (NRC) to M.E. Reddemann, Columbia Generating Station -

Issuance of Amendment [226] Re: Implementation of Power Range Neutron Monitoring/Average Power Range Monitor/Rod Block Monitor/Technical Specifications/Maximum Extended Load Line Limit Analysis (PRNM/ARTS/MELLLA) (TAC NO. ME7905), dated January 31, 2014, (ADAMS Accession Number ML13317B623).

2. GO2-13-075, Letter from B.J. Sawatzke, Columbia Generating Station to NRC Response to Request for Additional Information Regarding License Amendment Request to Implement PRNM/ARTS/MELLLA, dated May 9, 2013, (ADAMS Accession Number ML13141A581)

GO2-18-057 Page 10 of 10

3. TSTF-GG-05-01 Revision 1, Writers Guide for Plant-Specific Improved Technical Specifications.

4. Letter from M.C. Thadani (NRC) to M.E. Reddemann, Columbia Generating Station - Issuance of Amendment [223] Re: Deletion or Modifications of License Conditions That Have Been Completed or are No Longer in Effect (TAC NO.

ME5903), dated March 30, 2012, (ADAMS Accession Number ML120800078).

5. Letter from P.S. Tam (NRC) to T.J. OConnor, Nuclear Management Company, LLC, Monticello Nuclear Generating Plant - Issuance of Amendment Re:

Removal of the Table of Contents Out of the Technical Specifications (TAC NO.

MD6027), dated November 8, 2007 (ADAMS Accession Number ML072960159).

6. Letter from N. Kalyaman (NRC) to J.E. Venable, Entergy Operations, Inc.,

Waterford Steam Electric Station, Unit 3 - Issuance of Amendment Re:

Modification of Technical Specification (TS) 5.3.1, Fuel Assemblies, TS 5.6.1, Criticality, TS 6.9.1.11.1, Core Operating Limits Reports, and Deletion of TS Index, (TAC No. MC3584), dated May 9, 2005 (ADAMS Accession Number ML051290368).

7. Entergy Operations, Inc., letter to U.S. NRC, "Supplement to Amendment Request NPF-38-258 to Modify Technical Specification (TS) 5.3.1, Fuel Assemblies and TS 6.9.1.11.1, Core Operating Limits Report, Waterford Steam Electric Station, Unit 3," dated March 8, 2005.

8. Letter from J.S. Wiebe (NRC) to B.C. Hansen, Exelon Generation Company, LLC, Braidwood Station, Units 1 And 2, And Byron Station, Unit Nos. 1 And 2 -

Issuance of Amendments Regarding Request to Delete Obsolete License Conditions and Make Administrative Changes to Technical Specifications (CAC NOS. MF9338, MF9339, MF9340, AND MF9341), dated July 5, 2017. (ADAMS Accession Number ML17088A703).

GO2-18-057 Proposed Technical Specification Mark-up Pages

TABLE OF CONTENTS 1.0 USE AND APPLICATION 1.1 Definitions .............................................................................................................. 1.1-1 1.2 Logical Connectors ................................................................................................ 1.2-1 1.3 Completion Times ................................................................................................. 1.3-1 1.4 Frequency ............................................................................................................. 1.4-1 2.0 SAFETY LIMITS (SLs) 2.1 SLs ........................................................................................................................ 2.0-1 2.2 SL Violations ......................................................................................................... 2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY ..................... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY .................................... 3.0-4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)............................................................................ 3.1.1-1 3.1.2 Reactivity Anomalies .......................................................................................... 3.1.2-1 3.1.3 Control Rod OPERABILITY ................................................................................ 3.1.3-1 3.1.4 Control Rod Scram Times .................................................................................. 3.1.4-1 3.1.5 Control Rod Scram Accumulators ...................................................................... 3.1.5-1 3.1.6 Rod Pattern Control ............................................................................................ 3.1.6-1 3.1.7 Standby Liquid Control (SLC) System ................................................................ 3.1.7-1 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves ................................... 3.1.8-1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ............. 3.2.1-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) ................................................. 3.2.2-1 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) ................................................... 3.2.3-1 3.2.4 Average Power Range Monitor (APRM) Gain and Setpoint ............................... 3.2.4-1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation ......................................... 3.3.1.1-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation ............................................... 3.3.1.2-1 3.3.1.3 Oscillation Power Range Monitor (OPRM) Instrumentation ............................ 3.3.1.3-1 3.3.2.1 Control Rod Block Instrumentation .................................................................. 3.3.2.1-1 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation ............. 3.3.2.2-1 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation ............................................ 3.3.3.1-1 3.3.3.2 Remote Shutdown System .............................................................................. 3.3.3.2-1 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation ............... 3.3.4.1-1 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation .................................................................... 3.3.4.2-1 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation ............................ 3.3.5.1-1 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation..................... 3.3.5.2-1 Columbia Generating Station i Amendment 169,171 225

TABLE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.6.1 Primary Containment Isolation Instrumentation .............................................. 3.3.6.1-1 3.3.6.2 Secondary Containment Isolation Instrumentation ......................................... 3.3.6.2-1 3.3.7.1 Control Room Emergency Filtration (CREF) System Instrumentation ............ 3.3.7.1-1 3.3.8.1 Loss of Power (LOP) Instrumentation ............................................................. 3.3.8.1-1 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring ......................... 3.3.8.2-1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating ........................................................................... 3.4.1-1 3.4.2 Jet Pumps .......................................................................................................... 3.4.2-1 3.4.3 Safety/Relief Valves (SRVs) - 25% RTP ........................................................ 3.4.3-1 3.4.4 Safety/Relief Valves (SRVs) - < 25% RTP ........................................................ 3.4.4-1 3.4.5 RCS Operational LEAKAGE .............................................................................. 3.4.5-1 3.4.6 RCS Pressure Isolation Valve (PIV) Leakage .................................................... 3.4.6-1 3.4.7 RCS Leakage Detection Instrumentation ........................................................... 3.4.7-1 3.4.8 RCS Specific Activity .......................................................................................... 3.4.8-1 3.4.9 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown ..... 3.4.9-1 3.4.10 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown.. 3.4.10-1 3.4.11 RCS Pressure and Temperature (P/T) Limits .................................................. 3.4.11-1 3.4.12 Reactor Steam Dome Pressure ....................................................................... 3.4.12-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating .............................................................................................. 3.5.1-1 3.5.2 ECCS - Shutdown .............................................................................................. 3.5.2-1 3.5.3 RCIC System ..................................................................................................... 3.5.3-1 3.6 CONTAINMENT SYSTEMS 3.6.1.1 Primary Containment ...................................................................................... 3.6.1.1-1 3.6.1.2 Primary Containment Air Lock ........................................................................ 3.6.1.2-1 3.6.1.3 Primary Containment Isolation Valves (PCIVs) ............................................... 3.6.1.3-1 3.6.1.4 Drywell Air Temperature ................................................................................. 3.6.1.4-1 3.6.1.5 Residual Heat Removal (RHR) Drywell Spray ................................................ 3.6.1.5-1 3.6.1.6 Reactor Building-to-Suppression Chamber Vacuum Breakers ....................... 3.6.1.6-1 3.6.1.7 Suppression Chamber-to-Drywell Vacuum Breakers ...................................... 3.6.1.7-1 3.6.2.1 Suppression Pool Average Temperature ........................................................ 3.6.2.1-1 3.6.2.2 Suppression Pool Water Level ........................................................................ 3.6.2.2-1 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling ............................. 3.6.2.3-1 3.6.3.1 Deleted 3.6.3.2 Primary Containment Atmosphere Mixing System .......................................... 3.6.3.2-1 3.6.3.3 Primary Containment Oxygen Concentration .................................................. 3.6.3.3-1 3.6.4.1 Secondary Containment .................................................................................. 3.6.4.1-1 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) .......................................... 3.6.4.2-1 3.6.4.3 Standby Gas Treatment (SGT) System .......................................................... 3.6.4.3-1 Columbia Generating Station ii Amendment 169,199 225

TABLE OF CONTENTS 3.7 PLANT SYSTEMS 3.7.1 Standby Service Water (SW) System and Ultimate Heat Sink (UHS) ................ 3.7.1-1 3.7.2 High Pressure Core Spray (HPCS) Service Water (SW) System ...................... 3.7.2-1 3.7.3 Control Room Emergency Filtration (CREF) System ......................................... 3.7.3-1 3.7.4 Control Room Air Conditioning (AC) System...................................................... 3.7.4-1 3.7.5 Main Condenser Offgas ..................................................................................... 3.7.5-1 3.7.6 Main Turbine Bypass System ............................................................................. 3.7.6-1 3.7.7 Spent Fuel Storage Pool Water Level ................................................................ 3.7.7-1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating ..................................................................................... 3.8.1-1 3.8.2 AC Sources - Shutdown ..................................................................................... 3.8.2-1 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air .......................................................... 3.8.3-1 3.8.4 DC Sources - Operating .................................................................................... 3.8.4-1 3.8.5 DC Sources - Shutdown ..................................................................................... 3.8.5-1 3.8.6 Battery Parameters ............................................................................................ 3.8.6-1 3.8.7 Distribution Systems - Operating ........................................................................ 3.8.7-1 3.8.8 Distribution Systems - Shutdown........................................................................ 3.8.8-1 3.9 REFUELING OPERATIONS 3.9.1 Refueling Equipment Interlocks .......................................................................... 3.9.1-1 3.9.2 Refuel Position One-Rod-Out Interlock .............................................................. 3.9.2-1 3.9.3 Control Rod Position .......................................................................................... 3.9.3-1 3.9.4 Control Rod Position Indication .......................................................................... 3.9.4-1 3.9.5 Control Rod OPERABILITY - Refueling ............................................................. 3.9.5-1 3.9.6 Reactor Pressure Vessel (RPV) Water Level - Irradiated Fuel .......................... 3.9.6-1 3.9.7 Reactor Pressure Vessel (RPV) Water Level - New Fuel or Control Rods ........ 3.9.7-1 3.9.8 Residual Heat Removal (RHR) - High Water Level ............................................ 3.9.8-1 3.9.9 Residual Heat Removal (RHR) - Low Water Level............................................. 3.9.9-1 3.9.10 Decay Time ...................................................................................................... 3.9.10-1 3.10 SPECIAL OPERATIONS 3.10.1 Inservice Leak and Hydrostatic Testing Operation ........................................... 3.10.1-1 3.10.2 Reactor Mode Switch Interlock Testing ............................................................ 3.10.2-1 3.10.3 Single Control Rod Withdrawal - Hot Shutdown ............................................... 3.10.3-1 3.10.4 Single Control Rod Withdrawal - Cold Shutdown ............................................. 3.10.4-1 3.10.5 Single Control Rod Drive (CRD) Removal - Refueling ..................................... 3.10.5-1 3.10.6 Multiple Control Rod Withdrawal - Refueling .................................................... 3.10.6-1 3.10.7 Control Rod Testing - Operating....................................................................... 3.10.7-1 3.10.8 SHUTDOWN MARGIN (SDM) Test - Refueling .............................................. 3.10.8-1 4.0 DESIGN FEATURES 4.1 Site Location ......................................................................................................... 4.0-1 4.2 Reactor Core ......................................................................................................... 4.0-1 4.3 Fuel Storage .......................................................................................................... 4.0-2 Columbia Generating Station iii Amendment 199,204 225

TABLE OF CONTENTS 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility ........................................................................................................ 5.1-1 5.2 Organization .......................................................................................................... 5.2-1 5.3 Unit Staff Qualifications ......................................................................................... 5.3-1 5.4 Procedures ............................................................................................................ 5.4-1 5.5 Programs and Manuals ......................................................................................... 5.5-1 5.6 Reporting Requirements ....................................................................................... 5.6-1 5.7 High Radiation Area .............................................................................................. 5.7-1 Columbia Generating Station iv Amendment 149,169 225

Definitions 1.1 1.1 Definitions LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per unit length of RATE (LHGR) fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all FUNCTIONAL TEST required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MAXIMUM FRACTION OF The MFLPD shall be the largest value of the fraction of limiting LIMITING POWER DENSITY power density (FLPD) in the core. The FLPD shall be the (MFLPD) LHGR existing at a given location divided by the specified LHGR limit for that bundle type.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power ratio (CPR) that RATIO (MCPR) exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE - OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Chapter 14, Initial Test Program of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or Columbia Generating Station 1.1-4 Amendment No. 149,169 225

APRM Gain and Setpoint (Prior to Implementation of PRNM Upgrade) 3.2.4 3.2 POWER DISTRIBUTION LIMITS 3.2.4 Average Power Range Monitor (APRM) Gain and Setpoint LCO 3.2.4 a. MFLPD shall be less than or equal to Fraction of RTP (FRTP); or

b. Each required APRM Flow Biased Simulated Thermal Power - High Function Allowable Value shall be modified by greater than or equal to the ratio of FRTP and the MFLPD; or

c. Each required APRM gain shall be adjusted such that the APRM readings are 100% times MFLPD.

APPLICABILITY: THERMAL POWER 25% RTP prior to implementation of Power Range Neutron Monitor (PRNM) upgrade.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Satisfy the requirements of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months LCO not met. the LCO.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER to < 25% RTP.

Time not met.

Columbia Generating Station 3.2.4-1 Amendment No. 169 225 226

APRM Gain and Setpoint (Prior to Implementation of PRNM Upgrade) 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 -------------------------------NOTE------------------------------

Not required to be met if SR 3.2.4.2 is satisfied for LCO 3.2.4.b or LCO 3.2.4.c requirements.

Verify MFLPD is within limits. Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 25% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter SR 3.2.4.2 -------------------------------NOTE------------------------------

Not required to be met if SR 3.2.4.1 is satisfied for LCO 3.2.4.a requirements.

Verify each required: 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

a. APRM Flow Biased Simulated Thermal Power -

High Function Allowable Value is modified by greater than or equal to the ratio of FRTP and the MFLPD; or

b. APRM gain is adjusted such that the APRM reading is 100% times MFLPD.

Columbia Generating Station 3.2.4-2 Amendment No. 169 225 226

RPS Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.1-1 prior to implementation of Power Range Neutron Monitor (PRNM) upgrade.

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months channels inoperable.

OR A.2 Place associated trip 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months system in trip.

B. One or more Functions B.1 Place channel in one trip 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months with one or more system in trip.

required channels inoperable in both trip OR systems.

B.2 Place one trip system in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months trip.

C. One or more Functions C.1 Restore RPS trip capability. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with RPS trip capability not maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, Table 3.3.1.1-1 for the or C not met. channel.

Columbia Generating Station 3.3.1.1-1 Amendment No. 169 225 226

RPS Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. As required by Required E.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action D.1 and POWER to < 30% RTP.

referenced in Table 3.3.1.1-1.

F. As required by Required F.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully insert Immediately Action D.1 and all insertable control rods in referenced in core cells containing one or Table 3.3.1.1-1. more fuel assemblies.

SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Columbia Generating Station 3.3.1.1-2 Amendment No. 169 225 226

RPS Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.2 -------------------------------NOTE------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER 25% RTP.

Verify the absolute difference between the average 7 days power range monitor (APRM) channels and the calculated power 2% RTP plus any gain adjustment required by LCO 3.2.4, "Average Power Range Monitor (APRM) Gain and Setpoint," while operating at 25% RTP.

SR 3.3.1.1.3 -------------------------------NOTE------------------------------

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.4 Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.5 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels overlap. withdrawing SRMs from the fully inserted position SR 3.3.1.1.6 -------------------------------NOTE------------------------------

Only required to be met during entry into MODE 2 from MODE 1.

Verify the IRM and APRM channels overlap. 7 days SR 3.3.1.1.7 Calibrate the local power range monitors. 1130 MWD/T average core exposure Columbia Generating Station 3.3.1.1-3 Amendment No. 169 225 226

RPS Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1.9 ------------------------------NOTES-----------------------------

1. Neutron detectors are excluded.

2. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL CALIBRATION. 184 days SR 3.3.1.1.10 ------------------------------NOTES-----------------------------

1. Neutron detectors are excluded.

2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL CALIBRATION. 18 months for Functions 1 through 4, 6, 7, and 9 through 11 AND 24 months for Functions 5 and 8 SR 3.3.1.1.11 Verify the APRM Flow Biased Simulated Thermal 18 months Power - High Function time constant is 7 seconds.

SR 3.3.1.1.12 Verify Turbine Throttle Valve - Closure, and Turbine 18 months Governor Valve Fast Closure Trip Oil Pressure -

Low Functions are not bypassed when THERMAL POWER is 30% RTP.

SR 3.3.1.1.13 Perform CHANNEL FUNCTIONAL TEST. 24 months Columbia Generating Station 3.3.1.1-4 Amendment No. 179 225 226

RPS Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.14 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months SR 3.3.1.1.15 ------------------------------NOTES-----------------------------

1 Neutron detectors are excluded.

2. Channel sensors for Functions 3 and 4 are excluded.

3. For Function 5, "n" equals 4 channels for the purpose of determining the STAGGERED TEST BASIS Frequency.

Verify the RPS RESPONSE TIME is within limits. 24 months on a STAGGERED TEST BASIS Columbia Generating Station 3.3.1.1-5 Amendment No. 169 225 226

RPS Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. Intermediate Range Monitors

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 122/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.10 SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.1 122/125 SR 3.3.1.1.4 divisions of full SR 3.3.1.1.10 scale SR 3.3.1.1.14

b. Inop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14

2. Average Power Range Monitors

a. Neutron Flux - High, 2 2 G SR 3.3.1.1.1 20% RTP Setdown SR 3.3.1.1.3 SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.9 SR 3.3.1.1.14

b. Flow Biased Simulated 1 2 F SR 3.3.1.1.1 0.58 W + 62% RTP Thermal Power - High SR 3.3.1.1.2 and 114.9% RTP SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.14

c. Fixed Neutron Flux - 1 2 F SR 3.3.1.1.1 120% RTP High SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.9 SR 3.3.1.1.14 SR 3.3.1.1.15

d. Inop 1,2 2 G SR 3.3.1.1.7 NA SR 3.3.1.1.8 SR 3.3.1.1.14 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

Columbia Generating Station 3.3.1.1-6 Amendment No. 169 225 226

RPS Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

3. Reactor Vessel Steam 1,2 2 G SR 3.3.1.1.8 1079 psig Dome Pressure - High SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

4. Reactor Vessel Water Level 1,2 2 G SR 3.3.1.1.1 9.5 inches

- Low, Level 3 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

5. Main Steam Isolation Valve 1 8 F SR 3.3.1.1.8 12.5% closed

- Closure SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

6. Primary Containment 1,2 2 G SR 3.3.1.1.8 1.88 psig Pressure - High SR 3.3.1.1.10 SR 3.3.1.1.14

7. Scram Discharge Volume Water Level - High

a. Transmitter/Level 1,2 2 G SR 3.3.1.1.1 529 ft 9 inches Indicating Switch SR 3.3.1.1.8 elevation SR 3.3.1.1.10 SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.1 529 ft 9 inches SR 3.3.1.1.8 elevation SR 3.3.1.1.10 SR 3.3.1.1.14

b. Transmitter/Level 1,2 2 G SR 3.3.1.1.8 529 ft 9 inches Switch SR 3.3.1.1.10(d)(e) elevation SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.8 529 ft 9 inches SR 3.3.1.1.10(d)(e) elevation SR 3.3.1.1.14 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(d) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and the as-left tolerances are specified in the Licensee Controlled Specifications.

Columbia Generating Station 3.3.1.1-7 Amendment No. 225 226 232

RPS Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.1.1 Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

8. Turbine Throttle Valve - 30% RTP 4 E SR 3.3.1.1.8 7% closed Closure SR 3.3.1.1.10 SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15

9. Turbine Governor Valve 30% RTP 2 E SR 3.3.1.1.8 1000 psig Fast Closure, Trip Oil SR 3.3.1.1.10 Pressure - Low SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15

10. Reactor Mode Switch - 1,2 2 G SR 3.3.1.1.13 NA Shutdown Position SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.13 SR 3.3.1.1.14 NA

11. Manual Scram 1,2 2 G SR 3.3.1.1.4 NA SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.4 NA SR 3.3.1.1.14 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

Columbia Generating Station 3.3.1.1-8 Amendment No. 225 226 232

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.1-1 after implementation of Power Range Neutron Monitor (PRNM) upgrade.

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months channels inoperable.

OR

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

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

A.2 Place associated trip 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months system in trip.

NOTE-------------- B.1 Place channel in one trip 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Not applicable for system in trip.

Functions 2.a, 2.b, 2.c, 2.d, or 2.f. OR B.2 Place one trip system in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months B. One or more Functions trip.

with one or more required channels inoperable in both trip systems.

Columbia Generating Station 3.3.1.1-1 Amendment No. 169 225 226

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions C.1 Restore RPS trip capability. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with RPS trip capability not maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, Table 3.3.1.1-1 for the or C not met. channel.

E. As required by Required E.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action D.1 and POWER to < 29.5% RTP.

referenced in Table 3.3.1.1-1.

F. As required by Required F.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully insert Immediately Action D.1 and all insertable control rods in referenced in core cells containing one or Table 3.3.1.1-1. more fuel assemblies.

Columbia Generating Station 3.3.1.1-2 Amendment No. 169 225 226 241

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I. As required by Required I.1 Initiate alternate method to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and detect and suppress referenced in thermal hydraulic instability Table 3.3.1.1-1. oscillations.

AND

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

LCO 3.0.4 is not applicable.

I.2 Restore required channels 120 days to OPERABLE.

J. Required Action and J.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER to less than the Time of Condition I not value specified in the met. COLR.

SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.3.1.1-3 Amendment No. 169 225 226 238

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.2 -------------------------------NOTE------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER 25% RTP.

Verify the absolute difference between the average In accordance power range monitor (APRM) channels and the with the calculated power 2% RTP while operating at Surveillance Frequency 25% RTP.

Control Program SR 3.3.1.1.3 -------------------------------NOTE------------------------------

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.4 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.5 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels overlap. withdrawing SRMs from the fully inserted position SR 3.3.1.1.6 -------------------------------NOTE------------------------------

Only required to be met during entry into MODE 2 from MODE 1.

Verify the IRM and APRM channels overlap. In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.3.1.1-4 Amendment No. 169 225 226 238

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.7 Calibrate the local power range monitors. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.9 Deleted.

SR 3.3.1.1.10 ------------------------------NOTES---------------------------

1. Neutron detectors are excluded.

2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

3. For Functions 2.b and 2.f, the recirculation flow transmitters that feed the APRMs are included.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.11 Deleted.

SR 3.3.1.1.12 Verify Turbine Throttle Valve - Closure, and In accordance with Turbine Governor Valve Fast Closure Trip Oil the Surveillance Pressure - Low Functions are not bypassed when Frequency Control THERMAL POWER is 29.5% RTP. Program SR 3.3.1.1.13 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.3.1.1-5 Amendment No. 179 225 226 238 241

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.14 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.15 ------------------------------NOTES-----------------------------

1. Neutron detectors are excluded.

2. Channel sensors for Functions 3 and 4 are excluded.

Verify the RPS RESPONSE TIME is within limits. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.16 ------------------------------NOTES-----------------------------

1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

2. For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow transmitters.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.17 Verify the OPRM is not bypassed when APRM In accordance Simulated Thermal Power is greater than or equal to with the the value specified in the COLR and recirculation Surveillance drive flow is less than the value specified in the Frequency COLR. Control Program Columbia Generating Station 3.3.1.1-6 Amendment No. - 169 -

225 - -

226 238

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 Table 3.3.1.1-1 (page 1 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. Intermediate Range Monitors

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 122/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.10 SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.1 122/125 SR 3.3.1.1.4 divisions of full SR 3.3.1.1.10 scale SR 3.3.1.1.14

b. Inop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14

2. Average Power Range Monitors

a. Neutron Flux - High 2 3(b) G SR 3.3.1.1.1 20% RTP (Setdown) SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.10(d),(e)

SR 3.3.1.1.16

b. Simulated Thermal 1 3(b) F SR 3.3.1.1.1 0.62W + 62.9% RTP Power - High SR 3.3.1.1.2 and 114.9% RTP(c)

SR 3.3.1.1.7 SR 3.3.1.1.10(d),(e)

SR 3.3.1.1.16 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Each APRM/OPRM channel provides inputs to both trip systems.

(c) 0.62W + 59.8% RTP and 114.9% RTP when reset for single loop operation per LCO 3.4.1, Recirculation Loops Operating.

(d) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and as-left tolerances are specified in the Licensee Controlled Specifications.

Columbia Generating Station 3.3.1.1-7 Amendment No. 169 225 226 241

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 Table 3.3.1.1-1 (page 2 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

2. Average Power Range Monitors

c. Neutron Flux - High 1 3(b) F SR 3.3.1.1.1 120% RTP SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.10(d),(e)

SR 3.3.1.1.16

d. Inop 1,2 3(b) G SR 3.3.1.1.16 NA

e. 2-Out-of-4 Voter 1,2 2 G SR 3.3.1.1.1 NA SR 3.3.1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.16

f. OPRM Upscale (f) 3(b) I SR 3.3.1.1.1 NA(g)

SR 3.3.1.1.7 SR 3.3.1.1.10(d),(e)

SR 3.3.1.1.16 SR 3.3.1.1.17 (b) Each APRM/OPRM channel provides inputs to both trip systems.

(d) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and as-left tolerances are specified in the Licensee Controlled Specifications.

(f) THERMAL POWER greater than or equal to the value specified in the COLR.

(g) The OPRM Upscale does not have an Allowable Value. The Period Based Detection Algorithm (PBDA) trip setpoints are specified in the COLR.

Columbia Generating Station 3.3.1.1-8 Amendment No. 169 225 226

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 Table 3.3.1.1-1 (page 3 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

3. Reactor Vessel Steam 1,2 2 G SR 3.3.1.1.8 1079 psig Dome Pressure - High SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

4. Reactor Vessel Water 1,2 2 G SR 3.3.1.1.1 9.5 inches Level - Low, Level 3 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

5. Main Steam Isolation Valve 1 8 F SR 3.3.1.1.8 12.5% closed

- Closure SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

6. Primary Containment 1,2 2 G SR 3.3.1.1.8 1.88 psig Pressure - High SR 3.3.1.1.10 SR 3.3.1.1.14

7. Scram Discharge Volume Water Level - High

a. Transmitter/Level 1,2 2 G SR 3.3.1.1.1 529 ft 9 inches Indicating Switch SR 3.3.1.1.8 elevation SR 3.3.1.1.10 SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.1 529 ft 9 inches SR 3.3.1.1.8 elevation SR 3.3.1.1.10 SR 3.3.1.1.14

b. Transmitter/Level 1,2 2 G SR 3.3.1.1.8 529 ft 9 inches Switch SR 3.3.1.1.10(d)(e) elevation SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.8 529 ft 9 inches SR 3.3.1.1.10(d)(e) elevation SR 3.3.1.1.14 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(d) If the as-found channel setpoint is outside its predefinded as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and the as-left tolerances are specified in the Licensee Controlled Specifications.

Columbia Generating Station 3.3.1.1-9 Amendment No. 225 226 232

RPS Instrumentation (After Implementation of PRNM Upgrade) 3.3.1.1 Table 3.3.1.1-1 (page 4 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

8. Turbine Throttle Valve - 29.5% RTP 4 E SR 3.3.1.1.8 7% closed Closure SR 3.3.1.1.10 SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15

9. Turbine Governor Valve 29.5% RTP 2 E SR 3.3.1.1.8 1000 psig Fast Closure, Trip Oil SR 3.3.1.1.10 Pressure - Low SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15

10. Reactor Mode Switch - 1,2 2 G SR 3.3.1.1.13 NA Shutdown Position SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.13 SR 3.3.1.1.14 NA

11. Manual Scram 1,2 2 G SR 3.3.1.1.4 NA SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.4 NA SR 3.3.1.1.14 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

Columbia Generating Station 3.3.1.1-10 Amendment No. 225 226 232 241

OPRM Instrumentation OPRM Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.1.3 3.3 INSTRUMENTATION 3.3.1.3 Oscillation Power Range Monitor (OPRM) Instrumentation LCO 3.3.1.3 Four channels of the OPRM instrumentation shall be OPERABLE within the limits as specified in the COLR.

APPLICABILITY: THERMAL POWER 25% RTP prior to implementation of Power Range Neutron Monitor (PRNM) upgrade.

ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 30 days channels inoperable.

OR A.2 Place associated RPS trip 30 days system in trip.

OR A.3 Initiate alternate method to 30 days detect and suppress thermal hydraulic instability oscillations.

B. OPRM trip capability not B.1 Initiate alternate method to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months maintained. detect and suppress thermal hydraulic instability oscillations.

C. Required Action and C.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER < 25% RTP.

Time not met.

Columbia Generating Station 3.3.1.3-1 Amendment No. 171 225 226

OPRM Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.1.3 SURVEILLANCE REQUIREMENTS

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

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , provided the OPRM System maintains trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.3.1 Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1.3.2 Calibrate the local power range monitors. 1130 MWD/T average core exposure SR 3.3.1.3.3 -------------------------------NOTE------------------------------

Neutron detectors are excluded.

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.3.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months SR 3.3.1.3.5 Verify OPRM is not bypassed when THERMAL 24 months POWER is 30% RTP and core flow 60% rated core flow.

SR 3.3.1.3.6 -----------------------------NOTE--------------------------------

Neutron detectors are excluded.

Verify the RPS RESPONSE TIME is within limits. 24 months on a STAGGERED TEST BASIS Columbia Generating Station 3.3.1.3-2 Amendment No. 171 225 226

Control Rod Block Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.2.1 3.3 INSTRUMENTATION 3.3.2.1 Control Rod Block Instrumentation LCO 3.3.2.1 The control rod block instrumentation for each Function in Table 3.3.2.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.2.1-1 prior to implementation of Power Range Neutron Monitor (PRNM) upgrade.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One rod block monitor A.1 Restore RBM channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (RBM) channel OPERABLE status.

inoperable.

B. Required Action and B.1 Place one RBM channel in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated Completion trip.

Time of Condition A not met.

OR Two RBM channels inoperable.

C. Rod worth minimizer C.1 Suspend control rod Immediately (RWM) inoperable movement except by during reactor startup. scram.

OR C.2.1.1 Verify 12 rods withdrawn. Immediately OR Columbia Generating Station 3.3.2.1-1 Amendment No. 169 225 226

Control Rod Block Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.2.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2.1.2 Verify by administrative Immediately methods that startup with RWM inoperable has not been performed in the last calendar year.

AND C.2.2 Verify movement of control During control rod rods is in compliance with movement banked position withdrawal sequence (BPWS) by a second licensed operator or other qualified member of the technical staff.

D. RWM inoperable during D.1 Verify movement of control During control rod reactor shutdown. rods is in compliance with movement BPWS by a second licensed operator or other qualified member of the technical staff.

E. One or more Reactor E.1 Suspend control rod Immediately Mode Switch - Shutdown withdrawal.

Position channels inoperable. AND E.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies.

Columbia Generating Station 3.3.2.1-2 Amendment No. 169 225 226

Control Rod Block Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.2.1 SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.

2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains control rod block capability.

SURVEILLANCE FREQUENCY SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.2.1.2 -------------------------------NOTE------------------------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn at 10% RTP in MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.2.1.3 -------------------------------NOTE------------------------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is 10% RTP in MODE 1.

Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.2.1.4 -------------------------------NOTE------------------------------

Neutron detectors are excluded.

Verify the RBM is not bypassed: 92 days

a. When THERMAL POWER is 30% RTP; and

b. When a peripheral control rod is not selected.

Columbia Generating Station 3.3.2.1-3 Amendment No. 169 225 226

Control Rod Block Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.2.1.5 -------------------------------NOTE------------------------------

Neutron detectors are excluded.

Perform CHANNEL CALIBRATION. 92 days SR 3.3.2.1.6 Verify the RWM is not bypassed when THERMAL 24 months POWER is 10% RTP.

SR 3.3.2.1.7 -------------------------------NOTE------------------------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after reactor mode switch is in the shutdown position.

Perform CHANNEL FUNCTIONAL TEST. 24 months SR 3.3.2.1.8 Verify control rod sequences input to the RWM are Prior to declaring in conformance with BPWS. RWM OPERABLE following loading of sequence into RWM Columbia Generating Station 3.3.2.1-4 Amendment No. 179 225 226

Control Rod Block Instrumentation (Prior to Implementation of PRNM Upgrade) 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE

1. Rod Block Monitor

a. Upscale (a) 2 SR 3.3.2.1.1 0.58W + 51 %

SR 3.3.2.1.4 RTP SR 3.3.2.1.5

b. Inop (a) 2 SR 3.3.2.1.1 NA SR 3.3.2.1.4

c. Downscale (a) 2 SR 3.3.2.1.1 3% RTP SR 3.3.2.1.4 SR 3.3.2.1.5

2. Rod Worth Minimizer 1(b), 2(b) 1 SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.6 SR 3.3.2.1.8

3. Reactor Mode Switch - Shutdown (c) 2 SR 3.3.2.1.7 NA Position (a) THERMAL POWER 30% RTP and no peripheral control rod selected.

(b) With THERMAL POWER 10% RTP.

(c) Reactor mode switch in the shutdown position.

Columbia Generating Station 3.3.2.1-5 Amendment No. 169 225 226

Control Rod Block Instrumentation (Prior to Implementation of PRNM Upgrade) I 3.3.2.1 (This page intentionally blank)

Columbia Generating Station 3.3.2.1-6 Amendment No. 226

Control Rod Block Instrumentation (After Implementation of PRNM Upgrade) 3.3.2.1 3.3 INSTRUMENTATION 3.3.2.1 Control Rod Block Instrumentation LCO 3.3.2.1 The control rod block instrumentation for each Function in Table 3.3.2.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.2.1-1. after implementation of Power Range Neutron Monitor (PRNM) upgrade.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One rod block monitor A.1 Restore RBM channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (RBM) channel OPERABLE status.

inoperable.

B. Required Action and B.1 Place one RBM channel in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated Completion trip.

Time of Condition A not met.

OR Two RBM channels inoperable.

C. Rod worth minimizer C.1 Suspend control rod Immediately (RWM) inoperable movement except by during reactor startup. scram.

OR C.2.1.1 Verify 12 rods withdrawn. Immediately OR Columbia Generating Station 3.3.2.1-1 Amendment No. 169 225 226

Control Rod Block Instrumentation (After Implementation of PRNM Upgrade) 3.3.2.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2.1.2 Verify by administrative Immediately methods that startup with RWM inoperable has not been performed in the last calendar year.

AND C.2.2 Verify movement of control During control rod rods is in compliance with movement banked position withdrawal sequence (BPWS) by a second licensed operator or other qualified member of the technical staff.

D. RWM inoperable during D.1 Verify movement of control During control rod reactor shutdown. rods is in compliance with movement BPWS by a second licensed operator or other qualified member of the technical staff.

E. One or more Reactor E.1 Suspend control rod Immediately Mode Switch - Shutdown withdrawal.

Position channels inoperable. AND E.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies.

Columbia Generating Station 3.3.2.1-2 Amendment No. 169 225 226

Control Rod Block Instrumentation (After Implementation of PRNM Upgrade) 3.3.2.1 SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.

2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains control rod block capability.

SURVEILLANCE FREQUENCY SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.2 -------------------------------NOTE------------------------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after any control rod is withdrawn at 10% RTP in MODE 2.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.3 -------------------------------NOTE------------------------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after THERMAL POWER is 10% RTP in MODE 1.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.3.2.1-3 Amendment No. 169 225 226 238

Control Rod Block Instrumentation (After Implementation of PRNM Upgrade) 3.3.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.2.1.4 -------------------------------NOTE------------------------------

Neutron detectors are excluded.

Verify the RBM is not bypassed: In accordance with the

a. Low Power Range - Upscale Function is not Surveillance bypassed when APRM Simulated Thermal Frequency Power is 28% and < 63% RTP and peripheral Control Program control rod is not selected.

b. Intermediate Power Range - Upscale Function is not bypassed when APRM Simulated Thermal Power is 63% and < 83% RTP and peripheral control rod is not selected.

c. High Power Range - Upscale Function is not bypassed when APRM Simulated Thermal Power is 83% and peripheral control rod is not selected.

SR 3.3.2.1.5 -------------------------------NOTE------------------------------

Neutron detectors are excluded.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.6 Verify the RWM is not bypassed when THERMAL In accordance POWER is 10% RTP. with the Surveillance Frequency Control Program Columbia Generating Station 3.3.2.1-4 Amendment No. 179 225 226 238

Control Rod Block Instrumentation (After Implementation of PRNM Upgrade) 3.3.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.2.1.7 -------------------------------NOTE------------------------------

Not required to be performed until 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after reactor mode switch is in the shutdown position.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.2.1.8 Verify control rod sequences input to the RWM are Prior to declaring in conformance with BPWS. RWM OPERABLE following loading of sequence into RWM Columbia Generating Station 3.3.2.1-5 Amendment No. 226 238

Control Rod Block Instrumentation (After Implementation of PRNM Upgrade) 3.3.2.1 Table 3.3.2.1-1 (page 1 of 2)

Control Rod Block Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE

1. Rod Block Monitor

a. Low Power Range - Upscale (a) 2 SR 3.3.2.1.1 (f)

SR 3.3.2.1.4 SR 3.3.2.1.5(d),(e)

b. Intermediate Power Range - (b) 2 SR 3.3.2.1.1 (f)

Upscale SR 3.3.2.1.4 SR 3.3.2.1.5(d),(e)

c. High Power Range - Upscale (c) 2 SR 3.3.2.1.1 (f)

SR 3.3.2.1.4 SR 3.3.2.1.5(d),(e)

d. Inop (a),(b),(c) 2 SR 3.3.2.1.1 NA (a) APRM Simulated Thermal Power is 28% and < 63% RTP and MCPR is less than the limit specified in the COLR and no peripheral control rod selected.

(b) APRM Simulated Thermal Power is 63% and < 83% RTP and MCPR is less than the limit specified in the COLR and no peripheral control rod selected.

(c) APRM Simulated Thermal Power is 83% and MCPR is less than the limit specified in the COLR and no peripheral control rod selected.

(d) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and as-left tolerances are specified in the Licensee Controlled Specifications.

(f) Allowable Value Specified in the COLR.

Columbia Generating Station 3.3.2.1-6 Amendment No. 226 238

Control Rod Block Instrumentation (After Implementation of PRNM Upgrade) 3.3.2.1 Table 3.3.2.1-1 (page 2 of 2)

Control Rod Block Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE

2. Rod Worth Minimizer 1(g), 2(g) 1 SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.6 SR 3.3.2.1.8

3. Reactor Mode Switch - Shutdown (h) 2 SR 3.3.2.1.7 NA Position (g) With THERMAL POWER 10% RTP.

(h) Reactor mode switch in the shutdown position.

Columbia Generating Station 3.3.2.1-7 Amendment No. 238

Recirculation Loops Operating (Prior to Implementation of PRNM Upgrade) 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation.

OR One recirculation loop shall be in operation provided that the following limits are applied when the associated LCO is applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR; and

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR.

APPLICABILITY: MODES 1 and 2 prior to implementation of Power Range Neutron Monitor (PRNM) upgrade.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Recirculation loop flow A.1 Declare the recirculation 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months mismatch not within loop with lower flow to be limits. "not in operation."

B. Requirements of the B.1 Satisfy the requirements of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LCO not met for reasons the LCO.

other than Condition A.

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

OR No recirculation loops in operation.

Columbia Generating Station 3.4.1-1 Amendment No. 205 225 226

Recirculation Loops Operating (Prior to Implementation of PRNM Upgrade) 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 -------------------------------NOTE------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recirculation loops are in operation.

Verify recirculation loop drive flow mismatch with 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months both recirculation loops in operation is:

a. 10% of rated recirculation loop drive flow when operating at < 70% of rated core flow; and

b. 5% of rated recirculation loop drive flow when operating at 70% of rated core flow.

Columbia Generating Station 3.4.1-2 Amendment No. 171 225 226

Recirculation Loops Operating (After Implementation of PRNM Upgrade) 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation.

OR One recirculation loop shall be in operation provided that the following limits are applied when the associated LCO is applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR;

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; and

c. LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation, Function 2.b (Average Power Range Monitors, Simulated Thermal Power - High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.

APPLICABILITY: MODES 1 and 2 after implementation of Power Range Neutron Monitor (PRNM) upgrade.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Recirculation loop flow A.1 Declare the recirculation 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months mismatch not within loop with lower flow to be limits. "not in operation."

B. Requirements of the B.1 Satisfy the requirements of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LCO not met for reasons the LCO.

other than Condition A.

Columbia Generating Station 3.4.1-1 Amendment No. 205 225 226

Recirculation Loops Operating (After Implementation of PRNM Upgrade) 3.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A or B not met.

OR No recirculation loops in operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 -------------------------------NOTE------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recirculation loops are in operation.

Verify recirculation loop drive flow mismatch with In accordance both recirculation loops in operation is: with the Surveillance

a. 10% of rated recirculation loop drive flow Frequency when operating at < 70% of rated core flow; and Control Program

b. 5% of rated recirculation loop drive flow when operating at 70% of rated core flow.

Columbia Generating Station 3.4.1-2 Amendment No. 171 225 226 238

SDM Test - Refueling (Prior to Implementation of PRNM Upgrade) 3.10.8 3.10 SPECIAL OPERATIONS 3.10.8 SHUTDOWN MARGIN (SDM) Test - Refueling LCO 3.10.8 The reactor mode switch position specified in Table 1.1-1 for MODE 5 may be changed to include the startup/hot standby position, and operation considered not to be in MODE 2, to allow SDM testing, provided the following requirements are met:

a. LCO 3.3.1.1, "Reactor Protection System Instrumentation," MODE 2 requirements for Functions 2.a and 2.d of Table 3.3.1.1-1;

b. 1. LCO 3.3.2.1, "Control Rod Block Instrumentation," MODE 2 requirements for Function 2 of Table 3.3.2.1-1, with banked position withdrawal sequence requirements of SR 3.3.2.1.8 changed to require the control rod sequence to conform to the SDM test sequence, OR

2. Conformance to the approved control rod sequence for the SDM test is verified by a second licensed operator or other qualified member of the technical staff;

c. Each withdrawn control rod shall be coupled to the associated control rod drive (CRD);

d. All control rod withdrawals during out of sequence control rod moves shall be made in notch out mode;

e. No other CORE ALTERATIONS are in progress; and

f. CRD charging water header pressure 940 psig.

APPLICABILITY: MODE 5 with the reactor mode switch in startup/hot standby position prior to implementation of Power Range Neutron Monitor (PRNM) upgrade.

Columbia Generating Station 3.10.8-1 Amendment No. - 169

--225

- 226

SDM Test - Refueling (Prior to Implementation of PRNM Upgrade) 3.10.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE------------ --------------------NOTE-------------------

Separate Condition entry Rod worth minimizer may be is allowed for each bypassed as allowed by control rod. LCO 3.3.2.1, "Control Rod Block

Instrumentation," if required, to allow insertion of inoperable control One or more control rod and continued operation.

rods not coupled to its ------------------------------------------------

associated CRD.

A.1 Fully insert inoperable 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months control rod.

AND A.2 Disarm the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CRD.

B. One or more of the B.1 Place the reactor mode Immediately above requirements not switch in the shutdown or met for reasons other refuel position.

than Condition A.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.8.1 Perform the MODE 2 applicable SRs for According to the LCO 3.3.1.1, Functions 2.a and 2.d of applicable SRs Table 3.3.1.1-1.

SR 3.10.8.2 -------------------------------NOTE------------------------------

Not required to be met if SR 3.10.8.3 satisfied.

Perform the MODE 2 applicable SRs for According to the LCO 3.3.2.1, Function 2 of Table 3.3.2.1-1. applicable SRs Columbia Generating Station 3.10.8-2 Amendment No. 169 225 226

SDM Test - Refueling (Prior to Implementation of PRNM Upgrade) 3.10.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.8.3 -------------------------------NOTE------------------------------

Not required to be met if SR 3.10.8.2 satisfied.

Verify movement of control rods is in compliance During control rod with the approved control rod sequence for the SDM movement test by a second licensed operator or other qualified member of the technical staff.

SR 3.10.8.4 Verify no other CORE ALTERATIONS are in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months progress.

SR 3.10.8.5 Verify each withdrawn control rod does not go to the Each time the withdrawn overtravel position. control rod is withdrawn to "full out" position AND Prior to satisfying LCO 3.10.8.c requirement after work on control rod or CRD System that could affect coupling SR 3.10.8.6 Verify CRD charging water header pressure 7 days 940 psig.

Columbia Generating Station 3.10.8-3 Amendment No. 169 225 226

SDM Test - Refueling (Prior to Implementation of PRNM Upgrade) 3.10.8 (This page intentionally blank)

Columbia Generating Station 3.10.8-4 Amendment No. 226

SDM Test - Refueling (After Implementation of PRNM Upgrade) 3.10.8 3.10 SPECIAL OPERATIONS 3.10.8 SHUTDOWN MARGIN (SDM) Test - Refueling LCO 3.10.8 The reactor mode switch position specified in Table 1.1-1 for MODE 5 may be changed to include the startup/hot standby position, and operation considered not to be in MODE 2, to allow SDM testing, provided the following requirements are met:

a. LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation,"

MODE 2 requirements for Functions 2.a, 2.d, and 2.e of Table 3.3.1.1-1;

b. 1. LCO 3.3.2.1, "Control Rod Block Instrumentation," MODE 2 requirements for Function 2 of Table 3.3.2.1-1, with banked position withdrawal sequence requirements of SR 3.3.2.1.8 changed to require the control rod sequence to conform to the SDM test sequence, OR

2. Conformance to the approved control rod sequence for the SDM test is verified by a second licensed operator or other qualified member of the technical staff;

c. Each withdrawn control rod shall be coupled to the associated control rod drive (CRD);

d. All control rod withdrawals during out of sequence control rod moves shall be made in notch out mode;

e. No other CORE ALTERATIONS are in progress; and

f. CRD charging water header pressure 940 psig.

APPLICABILITY:

1 MODE 5 with the reactor mode switch in startup/hot standby position after implementation of Power Range Neutron Monitor (PRNM) upgrade.

Columbia Generating Station 3.10.8-1 Amendment No. 226

SDM Test - Refueling (After Implementation of PRNM Upgrade) 3.10.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE------------ --------------------NOTE-------------------

Separate Condition entry Rod worth minimizer may be is allowed for each bypassed as allowed by control rod. LCO 3.3.2.1, "Control Rod Block

Instrumentation," if required, to allow insertion of inoperable control One or more control rod and continued operation.

rods not coupled to its ------------------------------------------------

associated CRD.

A.1 Fully insert inoperable 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months control rod.

AND A.2 Disarm the associated 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CRD.

B. One or more of the B.1 Place the reactor mode Immediately above requirements not switch in the shutdown or met for reasons other refuel position.

than Condition A.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.8.1 Perform the MODE 2 applicable SRs for According to the LCO 3.3.1.1, Functions 2.a, 2.d, and 2.e of applicable SRs Table 3.3.1.1-1.

SR 3.10.8.2 -------------------------------NOTE------------------------------

Not required to be met if SR 3.10.8.3 satisfied.

Perform the MODE 2 applicable SRs for According to the LCO 3.3.2.1, Function 2 of Table 3.3.2.1-1. applicable SRs Columbia Generating Station 3.10.8-2 Amendment No. 169 225 226

SDM Test - Refueling (After Implementation of PRNM Upgrade) 3.10.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.10.8.3 -------------------------------NOTE------------------------------

Not required to be met if SR 3.10.8.2 satisfied.

Verify movement of control rods is in compliance During control rod with the approved control rod sequence for the SDM movement test by a second licensed operator or other qualified member of the technical staff.

SR 3.10.8.4 Verify no other CORE ALTERATIONS are in In accordance progress. with the Surveillance Frequency Control Program SR 3.10.8.5 Verify each withdrawn control rod does not go to the Each time the withdrawn overtravel position. control rod is withdrawn to "full out" position AND Prior to satisfying LCO 3.10.8.c requirement after work on control rod or CRD System that could affect coupling SR 3.10.8.6 Verify CRD charging water header pressure In accordance 940 psig. with the Surveillance Frequency Control Program Columbia Generating Station 3.10.8-3 Amendment No. 169 225 226 238

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 CORE OPERATING 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 APLHGR for Specification 3.2.1;

2. The MCPR for Specification 3.2.2;

3. The LHGR for Specification 3.2.3;

4. Deleted;LCO 3.3.1.3, Oscillation Power Range Monitor (OPRM)

Instrumentation prior to implementation of Power Range Neutron Monitor (PRNM) upgrade;

5. The Oscillation Power Range Monitor (OPRM) Instrumentation for Specification 3.3.1.1 after implementation of PRNM upgrade; and

6. The Rod Block Monitor Instrumentation for Specification 3.3.2.1.after implementation of PRNM upgrade.

b. 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. XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company

2. XN-NF-85-67(P)(A), "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," Exxon Nuclear Company

3. EMF-85-74(P) Supplement 1(P)(A) and Supplement 2(P)(A),

"RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model,"

Siemens Power Corporation

4. ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation

5. XN-NF-80-19(P)(A) Volume 1, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis,"

Exxon Nuclear Company

6. XN-NF-80-19(P)(A) Volume 4, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company Columbia Generating Station 5.6-2 Amendment No. 190 225 226

Rod Pattern Control 3.1.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Place the reactor mode 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months switch in the shutdown position.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify all OPERABLE control rods comply with In accordance BPWS. with the Surveillance Frequency Control Program Columbia Generating Station 3.1.6-2 Amendment No. 149,169 225 238

Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.6.1.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.3 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.4 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.5 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.6.1.7 -------------------------------NOTE------------------------------

Channel sensors for Functions 1.a, 1.b, and 1.c are excluded.

Verify the ISOLATION SYSTEM RESPONSE TIME In accordance is within limits. with the Surveillance Frequency Control Program Columbia Generating Station 3.3.6.1-4 Amendment No. 150,169 225 238

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 3 of 6)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

3. RCIC System Isolation

e. RCIC Equipment 1, 2, 3 1 F SR 3.3.6.1.3 180F Room Area SR 3.3.6.1.4 Temperature - High SR 3.3.6.1.6

f. RCIC Equipment 1, 2, 3 1 F SR 3.3.6.1.3 60F Room Area SR 3.3.6.1.4 Differential SR 3.3.6.1.6 Temperature - High

g. Reactor Water 1, 2, 3 1 F SR 3.3.6.1.3 180F Cleanup (RWCU) SR 3.3.6.1.4 System SR 3.3.6.1.6 RWCU/RCIC Steam Line Routing Area Temperature - High

h. Manual Initiation 1, 2, 3 1(b) G SR 3.3.6.1.6 NA

4. RWCU System Isolation

a. Differential Flow - 1, 2, 3 1 F SR 3.3.6.1.1 67.4 gpm High SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.6

b. Differential Flow - 1, 2, 3 1 F SR 3.3.6.1.2 46.5 seconds Time Delay SR 3.3.6.1.5 SR 3.3.6.1.6

c. Blowdown Flow - 1, 2, 3 1 F SR 3.3.6.1.1 271.7 gpm High SR 3.3.6.1.2 SR 3.3.6.1.5 SR 3.3.6.1.6 SR 3.3.6.1.7

d. Heat Exchanger 1, 2, 3 1 F SR 3.3.6.1.3 160F Room Area SR 3.3.6.1.4 Temperature - High SR 3.3.6.1.6 (b) RCIC Manual Initiation only inputs into one of the two trip systems.

Columbia Generating Station 3.3.6.1-7 Amendment No. 169,172 225

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 4 of 6)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

4. RWCU System Isolation

e. Heat Exchanger 1, 2, 3 1 F SR 3.3.6.1.3 70F Room Area SR 3.3.6.1.4 Ventilation SR 3.3.6.1.6 Differential Temperature - High

f. Pump Room Area 1, 2, 3 1 per room F SR 3.3.6.1.3 180F Temperature - High SR 3.3.6.1.4 SR 3.3.6.1.6

g. Pump Room Area 1, 2, 3 1 per room F SR 3.3.6.1.3 100F Ventilation SR 3.3.6.1.4 Differential SR 3.3.6.1.6 Temperature - High

h. RWCU/RCIC Line 1, 2, 3 1 F SR 3.3.6.1.3 180F Routing Area SR 3.3.6.1.4 Temperature - High SR 3.3.6.1.6

i. RWCU Line Routing 1, 2, 3 1 per room F SR 3.3.6.1.3 Area Temperature - SR 3.3.6.1.4 High SR 3.3.6.1.6 Room 409, 509 175F Areas Room 408, 511 180F Areas

j. Reactor Vessel 1, 2, 3 2 F SR 3.3.6.1.2 -58 inches Water Level - Low SR 3.3.6.1.4 Low, Level 2 SR 3.3.6.1.6

k. Standby Liquid 1, 2, 3 2(c) I SR 3.3.6.1.6 NA Control (SLC)

System Initiation

l. Manual Initiation 1, 2, 3 2 G SR 3.3.6.1.6 NA (c) SLC System Initiation only inputs into one of the two trip systems.

Columbia Generating Station 3.3.6.1-8 Amendment No. 172,199 225

Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 5 of 6)

Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

5. Residual Heat Removal (RHR) Shutdown Cooling (SDC)

SystemRHR SDC System Isolation

a. Pump Room Area 3 1 per room F SR 3.3.6.1.3 150F Temperature - High SR 3.3.6.1.4 SR 3.3.6.1.6

b. Pump Room Area 3 1 per room F SR 3.3.6.1.3 70F Ventilation SR 3.3.6.1.4 Differential SR 3.3.6.1.6 Temperature - High

c. Heat Exchanger 3 1 per room F SR 3.3.6.1.3 Area Temperature - SR 3.3.6.1.4 High SR 3.3.6.1.6 Room 505 Area 140F Room 507 Area 160F Room 605 Area 150F Room 606 Area 140F

d. Reactor Vessel 3, 4, 5 2(d) J SR 3.3.6.1.1 9.5 inches Water Level - Low, SR 3.3.6.1.2 Level 3 SR 3.3.6.1.4 SR 3.3.6.1.6

e. Reactor Vessel 1, 2, 3 1 F SR 3.3.6.1.2 135 psig Pressure - High SR 3.3.6.1.4 SR 3.3.6.1.6

f. Manual Initiation 1, 2, 3 2 G SR 3.3.6.1.6 NA (d) Only one trip system required in MODES 4 and 5 with RHR Shutdown Cooling System integrity maintained.

Columbia Generating Station 3.3.6.1-9 Amendment No. 161,169 225

LOP Instrumentation 3.3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.8.1.2 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.8.1.3 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.8.1.4 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.3.8.1-3 Amendment No. 149,169 225 238

ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

ACTIONS

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

LCO 3.0.4.b is not applicable to High Pressure Core Spray (HPCS).

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.1 Restore low pressure 7 days(1) injection/spray ECCS injection/spray subsystem inoperable. subsystem to OPERABLE status.

B High Pressure Core B.1 Verify by administrative Immediately Spray (HPCS) System means RCIC System is inoperable. OPERABLE when RCIC System is required to be OPERABLE.

AND B.2 Restore HPCS System to 14 days OPERABLE status.

(1) The Completion Time that one train of RHR (RHR-A) can be inoperable as specified by Required Action A.1 may be extended beyond the 7 day completion time up to 7 days to support restoration of RHR-A following pump and motor replacement. This footnote 23:59 PST February 28, 2019.

Columbia Generating Station 3.5.1-1 Amendment No. 187 225 230 245

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days;

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:

1. For noble gases: less than or equal to a dose rate of 500 mrems/yr to the whole body and less than or equal to a dose rate of 3000 mrems/yr to the skin, and

2. For iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half lives > 8 days: less than or equal to a dose rate of 1500 mrems/yr to any organ;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190; and

k. Limitations on venting and purging of the primary containment through the Standby Gas Treatment System to maintain releases as low as reasonably achievable.; and

l. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.

Columbia Generating Station 5.5-3 Amendment 169,182 225

GO2-18-057 Proposed Technical Specification Clean Pages

Definitions 1.1 1.1 Definitions LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per unit length of RATE (LHGR) fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all FUNCTIONAL TEST required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power ratio (CPR) that RATIO (MCPR) exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE - OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attenant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Chapter 14, Initial Test Program of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or Columbia Generating Station 1.1-4 Amendment No. 149,169 225

RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.1-1 ACTIONS

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

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months channels inoperable.

OR

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

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

A.2 Place associated trip 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months system in trip.

NOTE-------------- B.1 Place channel in one trip 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Not applicable for system in trip.

Functions 2.a, 2.b, 2.c, 2.d, or 2.f. OR B.2 Place one trip system in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months B. One or more Functions trip.

with one or more required channels inoperable in both trip systems.

Columbia Generating Station 3.3.1.1-1 Amendment No. 169 225 226

RPS Instrumentation 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions C.1 Restore RPS trip capability. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with RPS trip capability not maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, Table 3.3.1.1-1 for the or C not met. channel.

E. As required by Required E.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action D.1 and POWER to < 29.5% RTP.

referenced in Table 3.3.1.1-1.

F. As required by Required F.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully insert Immediately Action D.1 and all insertable control rods in referenced in core cells containing one or Table 3.3.1.1-1. more fuel assemblies.

Columbia Generating Station 3.3.1.1-2 Amendment No. 169 225 226 241

RPS Instrumentation 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I. As required by Required I.1 Initiate alternate method to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and detect and suppress referenced in thermal hydraulic instability Table 3.3.1.1-1. oscillations.

AND

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

LCO 3.0.4 is not applicable.

I.2 Restore required channels 120 days to OPERABLE.

J. Required Action and J.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER to less than the Time of Condition I not value specified in the met. COLR.

SURVEILLANCE REQUIREMENTS

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

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.3.1.1-3 Amendment No. 169 225 226 238

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.2 -------------------------------NOTE------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER 25% RTP.

Verify the absolute difference between the average In accordance power range monitor (APRM) channels and the with the calculated power 2% RTP while operating at Surveillance Frequency 25% RTP.

Control Program SR 3.3.1.1.3 -------------------------------NOTE------------------------------

Not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.4 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.5 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels overlap. withdrawing SRMs from the fully inserted position SR 3.3.1.1.6 -------------------------------NOTE------------------------------

Only required to be met during entry into MODE 2 from MODE 1.

Verify the IRM and APRM channels overlap. In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.3.1.1-4 Amendment No. 169 225 226 238

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.7 Calibrate the local power range monitors. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.9 Deleted.

SR 3.3.1.1.10 ------------------------------NOTES---------------------------

1. Neutron detectors are excluded.

2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

3. For Functions 2.b and 2.f, the recirculation flow transmitters that feed the APRMs are included.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.11 Deleted.

SR 3.3.1.1.12 Verify Turbine Throttle Valve - Closure, and In accordance with Turbine Governor Valve Fast Closure Trip Oil the Surveillance Pressure - Low Functions are not bypassed when Frequency Control THERMAL POWER is 29.5% RTP. Program SR 3.3.1.1.13 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.3.1.1-5 Amendment No. 179 225 226 238 241

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.14 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.15 ------------------------------NOTES-----------------------------

1. Neutron detectors are excluded.

2. Channel sensors for Functions 3 and 4 are excluded.

Verify the RPS RESPONSE TIME is within limits. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.16 ------------------------------NOTES-----------------------------

1. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

2. For Functions 2.b and 2.f, the CHANNEL FUNCTIONAL TEST includes the recirculation flow input processing, excluding the flow transmitters.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.17 Verify the OPRM is not bypassed when APRM In accordance Simulated Thermal Power is greater than or equal to with the the value specified in the COLR and recirculation Surveillance drive flow is less than the value specified in the Frequency COLR. Control Program Columbia Generating Station 3.3.1.1-6 Amendment No. - 169 -225 - -

226 238

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. Intermediate Range Monitors

a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 122/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.10 SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.1 122/125 SR 3.3.1.1.4 divisions of full SR 3.3.1.1.10 scale SR 3.3.1.1.14

b. Inop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14

2. Average Power Range Monitors

a. Neutron Flux - High 2 3(b) G SR 3.3.1.1.1 20% RTP (Setdown) SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.10(d),(e)

SR 3.3.1.1.16

b. Simulated Thermal 1 3(b) F SR 3.3.1.1.1 0.62W + 62.9% RTP Power - High SR 3.3.1.1.2 and 114.9% RTP(c)

SR 3.3.1.1.7 SR 3.3.1.1.10(d),(e)

SR 3.3.1.1.16 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Each APRM/OPRM channel provides inputs to both trip systems.

(c) 0.62W + 59.8% RTP and 114.9% RTP when reset for single loop operation per LCO 3.4.1, Recirculation Loops Operating.

(d) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and as-left tolerances are specified in the Licensee Controlled Specifications.

Columbia Generating Station 3.3.1.1-7 Amendment No. 169 225 226 241

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

2. Average Power Range Monitors

c. Neutron Flux - High 1 3(b) F SR 3.3.1.1.1 120% RTP SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.10(d),(e)

SR 3.3.1.1.16

d. Inop 1,2 3(b) G SR 3.3.1.1.16 NA

e. 2-Out-of-4 Voter 1,2 2 G SR 3.3.1.1.1 NA SR 3.3.1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.16

f. OPRM Upscale (f) 3(b) I SR 3.3.1.1.1 NA(g)

SR 3.3.1.1.7 SR 3.3.1.1.10(d),(e)

SR 3.3.1.1.16 SR 3.3.1.1.17 (b) Each APRM/OPRM channel provides inputs to both trip systems.

(d) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and as-left tolerances are specified in the Licensee Controlled Specifications.

(f) THERMAL POWER greater than or equal to the value specified in the COLR.

(g) The OPRM Upscale does not have an Allowable Value. The Period Based Detection Algorithm (PBDA) trip setpoints are specified in the COLR.

Columbia Generating Station 3.3.1.1-8 Amendment No. 169 225 226

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

3. Reactor Vessel Steam 1,2 2 G SR 3.3.1.1.8 1079 psig Dome Pressure - High SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

4. Reactor Vessel Water 1,2 2 G SR 3.3.1.1.1 9.5 inches Level - Low, Level 3 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

5. Main Steam Isolation Valve 1 8 F SR 3.3.1.1.8 12.5% closed

- Closure SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

6. Primary Containment 1,2 2 G SR 3.3.1.1.8 1.88 psig Pressure - High SR 3.3.1.1.10 SR 3.3.1.1.14

7. Scram Discharge Volume Water Level - High

a. Transmitter/Level 1,2 2 G SR 3.3.1.1.1 529 ft 9 inches Indicating Switch SR 3.3.1.1.8 elevation SR 3.3.1.1.10 SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.1 529 ft 9 inches SR 3.3.1.1.8 elevation SR 3.3.1.1.10 SR 3.3.1.1.14

b. Transmitter/Level 1,2 2 G SR 3.3.1.1.8 529 ft 9 inches Switch SR 3.3.1.1.10(d)(e) elevation SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.8 529 ft 9 inches SR 3.3.1.1.10(d)(e) elevation SR 3.3.1.1.14 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(d) If the as-found channel setpoint is outside its predefinded as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and the as-left tolerances are specified in the Licensee Controlled Specifications.

Columbia Generating Station 3.3.1.1-9 Amendment No. 225 226 232

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 4 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

8. Turbine Throttle Valve - 29.5% RTP 4 E SR 3.3.1.1.8 7% closed Closure SR 3.3.1.1.10 SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15

9. Turbine Governor Valve 29.5% RTP 2 E SR 3.3.1.1.8 1000 psig Fast Closure, Trip Oil SR 3.3.1.1.10 Pressure - Low SR 3.3.1.1.12 SR 3.3.1.1.14 SR 3.3.1.1.15

10. Reactor Mode Switch - 1,2 2 G SR 3.3.1.1.13 NA Shutdown Position SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.13 SR 3.3.1.1.14 NA

11. Manual Scram 1,2 2 G SR 3.3.1.1.4 NA SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.4 NA SR 3.3.1.1.14 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

Columbia Generating Station 3.3.1.1-10 Amendment No. 225 226 232 241

Control Rod Block Instrumentation 3.3.2.1 3.3 INSTRUMENTATION 3.3.2.1 Control Rod Block Instrumentation LCO 3.3.2.1 The control rod block instrumentation for each Function in Table 3.3.2.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.2.1-1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One rod block monitor A.1 Restore RBM channel to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (RBM) channel OPERABLE status.

inoperable.

B. Required Action and B.1 Place one RBM channel in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated Completion trip.

Time of Condition A not met.

OR Two RBM channels inoperable.

C. Rod worth minimizer C.1 Suspend control rod Immediately (RWM) inoperable movement except by during reactor startup. scram.

OR C.2.1.1 Verify 12 rods withdrawn. Immediately OR Columbia Generating Station 3.3.2.1-1 Amendment No. 169 225 226

Control Rod Block Instrumentation 3.3.2.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2.1.2 Verify by administrative Immediately methods that startup with RWM inoperable has not been performed in the last calendar year.

AND C.2.2 Verify movement of control During control rod rods is in compliance with movement banked position withdrawal sequence (BPWS) by a second licensed operator or other qualified member of the technical staff.

D. RWM inoperable during D.1 Verify movement of control During control rod reactor shutdown. rods is in compliance with movement BPWS by a second licensed operator or other qualified member of the technical staff.

E. One or more Reactor E.1 Suspend control rod Immediately Mode Switch - Shutdown withdrawal.

Position channels inoperable. AND E.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies.

Columbia Generating Station 3.3.2.1-2 Amendment No. 169 225 226

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS