GO2-23-006, License Amendment Request to Clean-Up Operating License and Appendix a Technical Specifications
| ML23121A294 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 05/01/2023 |
| From: | David Brown Energy Northwest |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| GO2-23-006 | |
| Download: ML23121A294 (1) | |
Text
David P. Brown Columbia Generating Station P.O. Box 968, PE23 Richland, WA 99352-0968 Ph. 509.377.8385 l F. 509.377.4150 dpbrown@energy-northwest.com GO2-23-006 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 LICENSE AMENDMENT REQUEST TO CLEAN-UP OPERATING LICENSE AND APPENDIX A TECHNICAL SPECIFICATIONS
Dear Sir or Madam:
Pursuant to 10 CFR 50.90, Energy Northwest hereby requests an amendment to revise the Columbia Generating Station Operating License (OL) NPF-21 and Appendix A Technical Specifications (TS). As part of this request, Energy Northwest is proposing several clean-up changes to the OL and TS, including editorial changes and the removal of obsolete TS information, spelling error corrections, and removal of obsolete OL conditions. None of the proposed changes result in changes to technical or operating requirements.
The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that these changes involve no significant hazards considerations. The bases for these determinations are included in Enclosure 1 of this submittal.
The proposed TS markup pages are included as Enclosure 2 to this submittal. Clean pages of the proposed TS changes are included as Enclosure 3 of this submittal.
Markups of the proposed OL changes are included as Enclosure 4 of this submittal.
Clean pages of the proposed OL changes are included as Enclosure 5 to this submittal.
This letter and its enclosures contain no regulatory commitments.
Approval of the proposed amendment is requested within one year of the date of the submittal. Once approved, the amendment shall be implemented within 90 days.
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May 1, 2023 ENERGY NORTHWEST
GO2-23-006 Page 2 In accordance with 10 CFR 50.91, Energy Northwest is notifying the State of Washington of this amendment request by transmitting a copy of this letter and enclosures to the designated State Official.
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 ______ day of ___________, 2023.
Respectfully, David P. Brown Site Vice President
Enclosures:
As stated cc:
NRC RIV Regional Administrator NRC NRR Project Manager NRC Senior Resident Inspector/988C CD Sonoda - BPA/1399 EFSEC@efsec.wa.gov - EFSEC E Fordham - WDOH R Brice - WDOH L Albin - WDOH
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#
GO2-23-006 Page 1 of 6 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 proposes several editorial changes to the OL and TS, including, but not limited to, removal of obsolete OL conditions, the removal of obsolete TS information, and spelling error corrections. 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:
Section 3.3.1.1 Reactor Protection System (RPS) Instrumentation Section 3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation Section 3.5.2 Reactor Pressure Vessel (RPV) Water Inventory Control Section 3.7.5 Main Condenser Offgas Section 5.2 Organization Section 5.3 Unit Staff Qualifications The specific conditions of the OL affected by the changes are:
Condition 2.C (36)
Core Plate Wedges Installation 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.
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, spelling, and formatting errors in various sections of TS. Energy Northwest is addressing the reasoning and descriptions for the proposed changes below.
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GO2-23-006 Page 2 of 6 2.2 Reason and Description for the Proposed Changes 2.2.1 Correction to Table 3.3.5.1-1 Note (e)
Revise Table 3.3.5.1-1 note (e) to delete reference to LCO 3.8.2, AC Sources -
Shutdown, as this note is no longer applicable in Mode 4 and 5. Changes made via the adoption of TSTF-542 (Reference 1) included revision from automatic initiation to manual initiation of the time delay relays for the start-up transformer; thus, the specified equipment no longer supports OPERABILITY of 230 kV offsite power circuits in Mode 4 and 5.
2.2.2 Correction of Typographical, Spelling, and Formatting Errors The correction of typographical, spelling, and formatting errors in the TS are detailed in Section 3.2 below. The corrections are supported by the guidance contained in the Writer's Guide for Plant-Specific Improved Technical Specifications (Reference 2).
2.2.3 Change in Title of Operations Manager to Operations Director Per a recent reorganization change, the title of Operations Manager was changed to Operations Director. This change needs to be reflected in Section 5.2.2 and 5.3.1 of the TS.
2.2.4 Removal of License Condition 2.C (36) - Core Plate Wedges The installation of the core plate wedges was completed, as reported in letter GO2 093, dated August 23, 2021, to the NRC (Reference 3). This report completes license condition 2.C (36), which states that the core plate wedges be installed on or before December 20, 2021.
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 Correction to Table 3.3.5.1-1 Note (e)
This change should have been included as part of the TS change to adopt TSTF-542, which was approved as Amendment 251 (Reference 1). All of the Mode 4 and 5 requirements were removed from TS 3.3.5.1, thus the reference to LCO 3.8.2 should have been removed. The relevant Mode 4 and 5 instrumentation requirements were relocated to TS 3.3.5.2. The LOCA time delay relays are not included in TS 3.3.5.2.
The ECCS initiation requirements in Mode 4 and 5 were revised from automatic initiation to manual initiation. As a result, the relays are not required to support the
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GO2-23-006 Page 3 of 6
operability of Start-Up Transformer in Mode 4 and 5. This was confirmed as part of the validation for the TSTF-542 amendment.
The corrections are detailed below.
- 1. TS page 3.3.5.1-10 note (e)
- 2. TS page 3.3.5.1-11 note (e) 3.2 Correction to Typographical, Spelling, and Formatting Errors The correction of typographical, spelling, and formatting errors in the TS are detailed below.
- 1. TS page 3.3.1.1-11 Note (d) the word predefined is spelled incorrectly
- 2. TS page 3.5.2-5 SR 3.5.2.7 alignment of note and text in table
- 3. TS page 3.7.5-2 contains a stray 5 in the word REQUIREMENTS 3.3 Change in Title of Operations Manager to Operations Director The title of Operations Manager was changed to Operation Director effective January 7, 2023. This change only affects job titles and does not change reporting structure.
The corrections in title names are detailed below.
- 1. TS page 5.2-2 (e) Operations Manager updated to Operations Director
- 2. TS page 5.3-1 (a) Operations Manager updated to Operations Director 3.4 Removal of License Condition 2.C (36) - Core Plate Wedges The condition was added to the License with NUREG-2123, Volume 1, Safety Evaluation Report Related to the License Renewal of Columbia Generating Station, published May 2012.
The installation of the core plate wedges was completed, as reported in letter GO2 093, dated August 23, 2021, to the NRC (Reference 3). This report completes license condition 2.C (36), which states that the core plate wedges be installed on or before December 20, 2021.
Completed one-time action license conditions do not need to be retained in the OL.
4.0 REGULATORY EVALUATION
The Columbia FSAR Section 1.8 provides detailed discussion of Columbias compliance with the applicable NRC regulatory guides.
The proposed OL and TS amendment is administrative in nature and:
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GO2-23-006 Page 4 of 6 x Does not result in any change in the qualifications of any component; and x 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 similar license amendment (No. 253, Reference 4) was issued to Columbia Generating Station on June 20, 2019, that allowed clean-up of the OL and TS, which included editorial changes and removal of obsolete information.
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.
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 OL or 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 FSAR. The proposed changes do not affect SSCs, operating procedures, or administrative controls that have the function of preventing or mitigating any of these accidents.
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GO2-23-006 Page 5 of 6 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 change does 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 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
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GO2-23-006 Page 6 of 6 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 L. John Klos (NRC) to Mr. Bradley J. Sawatzke (Energy Northwest),
Issuance of Amendment RE: Revision to Technical Specifications to Adopt TSTF-542, Revision 2, Reactor Pressure Vessel Water Inventory Control, (EPID L-2017-LLA-0361), dated October 30, 2018, ADAMS Accession No. ML18255A350.
- 2. TSTF-GG-05-01 Revision 1, Writers Guide for Plant-Specific Improved Technical Specifications.
- 3. Letter from A. Javorik (Energy Northwest) to U.S. Nuclear Regulatory Commission, Core Plate Wedge Installation Report Required by License Condition 2.C (36),
dated August 23, 2021, GO2-21-093, ADAMS Accession No. ML21235A454.
- 4. Letter from L. John Klos (NRC) to Mr. Bradley J. Sawatzke (Energy Northwest),
Issuance of Amendment No. 253 Re: Renewed Facility Operating License and Technical Specification Clean Up (EPID L-2018-LLA-0176), dated June 20, 2019, ADAMS Accession No. ML19063A579.
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GO2-23-006 Proposed Technical Specification Mark-up Pages
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RPS Instrumentation 3.3.1.1 Columbia Generating Station 3.3.1.1-11 Amendment No. 270 Table 3.3.1.1-1 (page 3 of 4)
Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE
- 3.
Reactor Vessel Steam Dome Pressure - High 1,2 2
G SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15 d 1079 psig
- 4.
Reactor Vessel Water Level - Low, Level 3 1,2 2
G SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15 t 9.5 inches
- 5.
- Closure 1
8 F
SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15 d 12.5% closed
- 6.
Primary Containment Pressure - High 1,2 2
G SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 d 1.88 psig
- 7.
Scram Discharge Volume Water Level - High
- a.
Transmitter/Level Indicating Switch 1,2 2
G SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 d 529 ft 9 inches elevation 5(a) 2 H
SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 d 529 ft 9 inches elevation
- b.
Transmitter/Level Switch 1,2 2
G SR 3.3.1.1.8 SR 3.3.1.1.10(d)(e)
SR 3.3.1.1.14 d 529 ft 9 inches elevation 5(a) 2 H
SR 3.3.1.1.8 SR 3.3.1.1.10(d)(e)
SR 3.3.1.1.14 d 529 ft 9 inches elevation (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 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.
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ECCS Instrumentation 3.3.5.1 Columbia Generating Station 3.3.5.1-10 Amendment No. 166, 169 225 238 251 270 Table 3.3.5.1-1 (page 1 of 5)
Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE
- 1. Low Pressure Coolant Injection-A (LPCI) and Low Pressure Core Spray (LPCS)
Subsystems
- a.
Reactor Vessel Water Level - Low Low Low, Level 1 1, 2, 3 2(b)
B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t -142.3 inches
- b.
Drywell Pressure -
High 1, 2, 3 2(b)
B SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 d 1.88 psig
- c.
LPCS Pump Start -
LOCA Time Delay Relay 1, 2, 3 1(e)
C SR 3.3.5.1.5 SR 3.3.5.1.6 t 8.53 seconds and d 10.64 seconds
- d.
LPCI Pump A Start -
LOCA Time Delay Relay 1, 2, 3 1(e)
C SR 3.3.5.1.5 SR 3.3.5.1.6 t 17.24 seconds and d 21.53 seconds
- e.
LPCI Pump A Start -
LOCA/LOOP Time Delay Relay 1, 2, 3 1
C SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.6 t 3.04 seconds and d 6.00 seconds
- f.
Reactor Vessel Pressure - Low (Injection Permissive) 1, 2, 3 1 per valve C
SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 448 psig and d 492 psig (a)
Deleted (b)
Also required to initiate the associated diesel generator (DG).
(e)
Also supports OPERABILITY of 230 kV offsite power circuit pursuant to LCO 3.8.1 and LCO 3.8.2.
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ECCS Instrumentation 3.3.5.1 Columbia Generating Station 3.3.5.1-11 Amendment No.238 270 Table 3.3.5.1-1 (page 2 of 5)
Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE 1.
LPCS Pump Discharge Flow -
Low (Minimum Flow) 1, 2, 3 1
E SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 668 gpm and d 1067 gpm h.
LPCI Pump A Discharge Flow -
Low (Minimum Flow) 1, 2, 3 1
E SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 605 gpm and d 984 gpm i.
Manual Initiation 1, 2, 3 2
C SR 3.3.5.1.6 NA 2.
LPCI B and LPCI C Subsystems a.
Reactor Vessel Water Level - Low Low Low, Level 1 1, 2, 3 2(b)
B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t -142.3 inches b.
Drywell Pressure -
High 1, 2, 3 2(b)
B SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 d 1.88 psig c.
LPCI Pump B Start -
LOCA Time Delay Relay 1, 2, 3 1(e)
C SR 3.3.5.1.5 SR 3.3.5.1.6 t 17.24 seconds and d 21.53 seconds d.
LPCI Pump C Start -
LOCA Time Delay Relay 1, 2, 3 1(e)
C SR 3.3.5.1.5 SR 3.3.5.1.6 t 8.53 seconds and d 10.64 seconds e.
LPCI Pump B Start -
LOCA/LOOP Time Delay Relay 1, 2, 3 1
C SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.6 t 3.04 seconds and d 6.00 seconds (a)
Deleted (b)
Also required to initiate the associated DG.
(e)
Also supports OPERABILITY of 230 kV offsite power circuit pursuant to LCO 3.8.1 and LCO 3.8.2.
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RPV Water Inventory Control 3.5.2 Columbia Generating Station 3.5.2-5 Amendment No. 251 264 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify, for the required ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.5
NOTE------------------------------
1.
Operation may be through the test return line.
2.
Credit may be taken for normal system operation to satisfy this SR.
Operate the required ECCS injection/spray subsystem for 10 minutes.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.6 Verify each valve credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated isolation signal.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.7
NOTE------------------------------
Vessel injection/spray may be excluded.
Verify the required ECCS injection/spray subsystem can be manually operated.
In accordance with the Surveillance Frequency Control Program
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Adjusted alignment of text in this SR
Main Condenser Offgas 3.7.5 Columbia Generating Station 3.7.5-2 Amendment No. 149,169 225 238 SURVEILLANCE REQUIR5EMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1
NOTE------------------------------
Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.
Verify the gross gamma activity rate of the noble gases is d 332 mCi/second after decay of 30 minutes.
In accordance with the Surveillance Frequency Control Program AND Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a t 50% increase in the nominal steady state fission gas release after factoring out increases due to changes in THERMAL POWER level
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Organization 5.2 Columbia Generation Station 5.2-2 Amendment 182,213 225 5.2 Organization 5.2.2 Unit Staff (continued)
F
An individual qualified to implement radiation protection procedures shall be
on site when fuel is in the reactor. The position may be vacant for not more
than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided
immediate action is taken to fill the required position.
G
Deleted.
H
The Operations ManagerDirector or Assistant Operations Manager shall
hold an SRO license.
I
An individual shall provide advisory technical support to the unit operations
shift crew in the areas of thermal hydraulics, reactor engineering, and plant
analysis with regard to the safe operation of the unit. This individual shall
meet the qualifications specified by the Commission Policy Statement on
Engineering Expertise on Shift.
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Unit Staff Qualifications 5.3 Columbia Generation Station 5.3-1 Amendment 169,182 225 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI/ANS N18.1-1971, for comparable positions described in the FSAR, except for:
- a.
The Operations ManagerDirector, who shall meet the requirements of ANSI/ANS N18.1-1971 with the exception that in lieu of meeting the stated ANSI/ANS requirement to hold a Senior Reactor Operator (SRO) license at the time of appointment to the position, the Operations ManagerDirector shall:
- 1.
Hold an SRO license at the time of appointment;
- 2.
Have held an SRO license; or
- 3.
Have been certified for equivalent SRO knowledge; and
- b.
The Radiation Protection Manager, who shall meet or exceed the qualifications of Regulatory Guide 1.8, Revision 1-R, May 1977.
5.3.2 For the purpose of 10 CFR 55.4, a licensed SRO and a licensed Reactor Operator (RO) are those individuals who, in addition to meeting the requirements of Specification 5.3.1, perform the functions described in 10 CFR 50.54(m).
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GO2-23-006 Proposed Technical Specification Clean Pages
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RPS Instrumentation 3.3.1.1 Columbia Generating Station 3.3.1.1-11 Amendment No. 270 Table 3.3.1.1-1 (page 3 of 4)
Reactor Protection System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION D.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE
- 3.
Reactor Vessel Steam Dome Pressure - High 1,2 2
G SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15 d 1079 psig
- 4.
Reactor Vessel Water Level - Low, Level 3 1,2 2
G SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15 t 9.5 inches
- 5.
- Closure 1
8 F
SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15 d 12.5% closed
- 6.
Primary Containment Pressure - High 1,2 2
G SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 d 1.88 psig
- 7.
Scram Discharge Volume Water Level - High
- a.
Transmitter/Level Indicating Switch 1,2 2
G SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 d 529 ft 9 inches elevation 5(a) 2 H
SR 3.3.1.1.1 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 d 529 ft 9 inches elevation
- b.
Transmitter/Level Switch 1,2 2
G SR 3.3.1.1.8 SR 3.3.1.1.10(d)(e)
SR 3.3.1.1.14 d 529 ft 9 inches elevation 5(a) 2 H
SR 3.3.1.1.8 SR 3.3.1.1.10(d)(e)
SR 3.3.1.1.14 d 529 ft 9 inches elevation (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.
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ECCS Instrumentation 3.3.5.1 Columbia Generating Station 3.3.5.1-10 Amendment No. 166, 169 225 238 251 270 Table 3.3.5.1-1 (page 1 of 5)
Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE
- 1. Low Pressure Coolant Injection-A (LPCI) and Low Pressure Core Spray (LPCS)
Subsystems
- a.
Reactor Vessel Water Level - Low Low Low, Level 1 1, 2, 3 2(b)
B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t -142.3 inches
- b.
Drywell Pressure -
High 1, 2, 3 2(b)
B SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 d 1.88 psig
- c.
LPCS Pump Start -
LOCA Time Delay Relay 1, 2, 3 1(e)
C SR 3.3.5.1.5 SR 3.3.5.1.6 t 8.53 seconds and d 10.64 seconds
- d.
LPCI Pump A Start -
LOCA Time Delay Relay 1, 2, 3 1(e)
C SR 3.3.5.1.5 SR 3.3.5.1.6 t 17.24 seconds and d 21.53 seconds
- e.
LPCI Pump A Start -
LOCA/LOOP Time Delay Relay 1, 2, 3 1
C SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.6 t 3.04 seconds and d 6.00 seconds
- f.
Reactor Vessel Pressure - Low (Injection Permissive) 1, 2, 3 1 per valve C
SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 448 psig and d 492 psig (a)
Deleted (b)
Also required to initiate the associated diesel generator (DG).
(e)
Also supports OPERABILITY of 230 kV offsite power circuit pursuant to LCO 3.8.1.
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ECCS Instrumentation 3.3.5.1 Columbia Generating Station 3.3.5.1-11 Amendment No.238 270 Table 3.3.5.1-1 (page 2 of 5)
Emergency Core Cooling System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER FUNCTION CONDITIONS REFERENCED FROM REQUIRED ACTION A.1 SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE
- g.
LPCS Pump Discharge Flow -
Low (Minimum Flow) 1, 2, 3 1
E SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 668 gpm and d 1067 gpm
- h.
LPCI Pump A Discharge Flow -
Low (Minimum Flow) 1, 2, 3 1
E SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t 605 gpm and d 984 gpm
- i.
Manual Initiation 1, 2, 3 2
C SR 3.3.5.1.6 NA
- a.
Reactor Vessel Water Level - Low Low Low, Level 1 1, 2, 3 2(b)
B SR 3.3.5.1.1 SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 t -142.3 inches
- b.
Drywell Pressure -
High 1, 2, 3 2(b)
B SR 3.3.5.1.2 SR 3.3.5.1.4 SR 3.3.5.1.6 d 1.88 psig
- c.
LPCI Pump B Start -
LOCA Time Delay Relay 1, 2, 3 1(e)
C SR 3.3.5.1.5 SR 3.3.5.1.6 t 17.24 seconds and d 21.53 seconds
- d.
LPCI Pump C Start -
LOCA Time Delay Relay 1, 2, 3 1(e)
C SR 3.3.5.1.5 SR 3.3.5.1.6 t 8.53 seconds and d 10.64 seconds
- e.
LPCI Pump B Start -
LOCA/LOOP Time Delay Relay 1, 2, 3 1
C SR 3.3.5.1.2 SR 3.3.5.1.3 SR 3.3.5.1.6 t 3.04 seconds and d 6.00 seconds (a)
Deleted (b)
Also required to initiate the associated DG.
(e)
Also supports OPERABILITY of 230 kV offsite power circuit pursuant to LCO 3.8.1.
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RPV Water Inventory Control 3.5.2 Columbia Generating Station 3.5.2-5 Amendment No. 251 264 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify, for the required ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.5
NOTE------------------------------
- 1. Operation may be through the test return line.
- 2. Credit may be taken for normal system operation to satisfy this SR.
Operate the required ECCS injection/spray subsystem for 10 minutes.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.6 Verify each valve credited for automatically isolating a penetration flow path actuates to the isolation position on an actual or simulated isolation signal.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.7
NOTE------------------------------
Vessel injection/spray may be excluded.
Verify the required ECCS injection/spray subsystem can be manually operated.
In accordance with the Surveillance Frequency Control Program
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Main Condenser Offgas 3.7.5 Columbia Generating Station 3.7.5-2 Amendment No. 149,169 225 238 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1
NOTE------------------------------
Not required to be performed until 31 days after any main steam line not isolated and SJAE in operation.
Verify the gross gamma activity rate of the noble gases is d 332 mCi/second after decay of 30 minutes.
In accordance with the Surveillance Frequency Control Program AND Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a t 50% increase in the nominal steady state fission gas release after factoring out increases due to changes in THERMAL POWER level
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Organization 5.2 Columbia Generation Station 5.2-2 Amendment 182,213 225 5.2 Organization 5.2.2 Unit Staff (continued)
- c.
An individual qualified to implement radiation protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.
- d.
Deleted.
- e.
The Operations Director or Assistant Operations Manager shall hold an SRO license.
- f.
An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.
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Unit Staff Qualifications 5.3 Columbia Generation Station 5.3-1 Amendment 169,182 225 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI/ANS N18.1-1971, for comparable positions described in the FSAR, except for:
- a.
The Operations Director, who shall meet the requirements of ANSI/ANS N18.1-1971 with the exception that in lieu of meeting the stated ANSI/ANS requirement to hold a Senior Reactor Operator (SRO) license at the time of appointment to the position, the Operations Director shall:
- 1.
Hold an SRO license at the time of appointment;
- 2.
Have held an SRO license; or
- 3.
Have been certified for equivalent SRO knowledge; and
- b.
The Radiation Protection Manager, who shall meet or exceed the qualifications of Regulatory Guide 1.8, Revision 1-R, May 1977.
5.3.2 For the purpose of 10 CFR 55.4, a licensed SRO and a licensed Reactor Operator (RO) are those individuals who, in addition to meeting the requirements of Specification 5.3.1, perform the functions described in 10 CFR 50.54(m).
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GO2-23-006 Proposed Operating License Mark-up Pages
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Renewed License No. NPF-21 Amendment No. 225, 266 269 (34)
Deleted (35)
The licensee's FSAR, as updated with the license renewal FSAR supplement submitted pursuant to 10 CFR 54.21(d) and supplemented with Appendix A of NUREG-2123 with the exception of Commitments Nos. 55, 56, 57, and 71, and as revised pursuant to the criteria set forth in 10 CFR 50.59, describes certain future programs and activities to be completed before the period of extended operation. Energy Northwest shall complete these activities no later than July 20, 2023 and shall notify the NRC in writing when implementation of these activities is complete.
(36)
To prevent lateral motion of the core plate, the licensee shall install core plate wedges around the periphery of the core plate within the shroud on or before December 20, 2021. Upon completion of the core plate wedge installation, the licensee shall submit a written report to the NRC staff summarizing the results of the installation. The licensee shall also submit a written report regarding any corrective action taken related to core plate rim hold-down bolts or core plate wedges and the results of extent of condition reviews on or before December 20, 2021. Deleted (37) 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Energy Northwest is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSC) using:
Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic risk; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in License Amendment No. 269 dated December 15, 2022.
Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.
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GO2-23-006 Proposed Operating License Clean Pages
! "
Renewed License No. NPF-21 Amendment No. 225, 266 269 (34)
Deleted (35)
The licensee's FSAR, as updated with the license renewal FSAR supplement submitted pursuant to 10 CFR 54.21(d) and supplemented with Appendix A of NUREG-2123 with the exception of Commitments Nos. 55, 56, 57, and 71, and as revised pursuant to the criteria set forth in 10 CFR 50.59, describes certain future programs and activities to be completed before the period of extended operation. Energy Northwest shall complete these activities no later than July 20, 2023 and shall notify the NRC in writing when implementation of these activities is complete.
(36)
Deleted (37) 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors Energy Northwest is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSC) using:
Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic risk; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in License Amendment No. 269 dated December 15, 2022.
Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above.
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