ML25083A158
| ML25083A158 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/21/2025 |
| From: | David Brown Energy Northwest |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| G02-25-016 | |
| Download: ML25083A158 (1) | |
Text
David P. Brown Columbia Generating Station P.O. Box 968, PE23 Richland, WA 99352-0968 509.377.8385 dpbrown@energy-northwest.com GO2-25-016 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
COLUMBIA GENERATING STATION, DOCKET NO. 50-397 LICENSE AMENDMENT REQUEST TO REVISE COLUMBIA GENERATING STATION TECHNICAL SPECIFICATION 3.3.2.1, CONTROL ROD BLOCK INSTRUMENTATION
Dear Sir or Madam:
In accordance with 10 CFR 50.90, Energy Northwest requests an amendment to revise the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. NPF-21 for Columbia Generating Station (Columbia).
The proposed amendment would modify Columbia TS 3.3.2.1, Control Rod Block Instrumentation, Required Action C.2.1.2, which restricts reactor startup with an inoperable Rod Worth Minimizer to once per calendar year. The proposed change to Required Action C.2.1.2 is to allow additional reactor startups with a new action to verify FRQWUROrod coupling checks have been performed prior to a reactor restart.
The enclosure contains the results of analyses conducted in support of this application, specifically, an evaluation of the proposed change, which includes a detailed description, technical and regulatory evaluation supporting a no significant hazards consideration, and environmental consideration. Attachment 1 provides the existing TS pages marked up to show the proposed change. Attachment 2 provides revised (clean) TS pages. provides existing TS Bases pages marked to show the proposed change and is provided for information only.
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March 21, 2025 ENERGY NORTHWEST
GO2-25-016 Page 2 of 2 Approval of the proposed amendment is requested within one year from the date of this letter. Once approved, the amendment shall be implemented within 120 days.
There are no regulatory commitments made in this submittal.
In accordance with 10 CFR 50.91, a copy of this amendment request, with enclosures, is being provided to the designated Washington State Official.
If there are any questions or if additional information is needed, please contact Ms. T. M.
Collis at 509-377-8463.
I declare under penalty of perjury that the foregoing is true and correct.
Executed this ______ day of ___________ 2025.
Respectfully, David P. Brown Site Vice President
Enclosure:
Evaluation of Proposed Change : Proposed Columbia Technical Specification Change (Mark-Up) : Revised Columbia Technical Specification Pages : Proposed Columbia Technical Specification Bases Change (Mark-Up for Information Only) cc:
NRC RIV Regional Administrator NRC NRR Project Manager NRC Senior Resident Inspector/988C CD Sonoda - BPA/1399 EFSECutc.wa.gov - EFSEC E Fordham - WDOH R Brice - WDOH L Albin - WDOH
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GO2-25-016 Enclosure Page 1 of 11 EVALUATION OF PROPOSED CHANGE 1.0
SUMMARY
DESCRIPTION This evaluation supports an amendment request to the Technical Specifications (TS),
Appendix A of Renewed Facility Operating License No. NPF-21 for Columbia Generating Station (Columbia). The proposed amendment would modify Required Action C.2.1.2 from TS 3.3.2.1, Control Rod Block Instrumentation, which restricts reactor startup with an inoperable Rod Worth Minimizer (RWM) to once per calendar year.
2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The Plant Processing Computer (PPC) provides station personnel with real time and historical data used to analyze events and monitor plant performance. The RWM subsystem within the PPC functions as a backup to the Operator during movement of control rods during reactor startup, shutdown, and low power level (less than 10% rated thermal power) control rod sequences. The use of the RWM minimizes the consequences of a design basis control rod drop accident (RDA) by enforcing pre-established control rod sequences that are designed to maintain the worth of any control rod to acceptable levels as determined by the RDA analysis.
The RWM aids in enforcing compliance with the analyzed rod position sequence by initiating appropriate rod withdrawal and rod insertion block signals if the control rod patterns followed by the Operator are not consistent with the pre-established sequence. The pre-established rod position sequences follow the banked position withdrawal sequence (BPWS) analyzed in General Electric Topical Report NEDO-21231, Banked Position Withdrawal Sequence (BPWS), dated January 1977 (Reference 1). This sequence is stored in the PPC memory and is based on control rod withdrawal procedures designed to limit (and thereby minimize) individual control rod worths such that the peak fuel enthalpy would remain below the specific energy design limit of 280 calories per gram (cal/g) should an RDA occur.
Conformance to the design basis enthalpy value ensures that the offsite dose consequences due to an RDA will be within the guidelines of 10 CFR $FFLGHQW6RXUFH7HUP.
The RWM does not interfere with normal reactor operation, does not affect any normal instrumentation displays associated with the selection of a control rod, and does not have any plant control functions. The RWM will not function on a loss of offsite power and the failure of the RWM does not itself cause rod patterns to be established. The RWM function is automatically bypassed at power levels greater than 10% rated thermal power, as there are sufficient void concentrations to preclude a rod drop exceeding 280 cal/g peak fuel enthalpy. At power levels less than 10% rated thermal power, the RWM may be manually bypassed by specific procedural control and verification by a second licensed Operator (or qualified member of the technical staff) that the first Operator is performing control rod movements in accordance with the BPWS.
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GO2-25-016 Enclosure Page 2 of 11 2.2 Current Technical Specification Requirements For continued rod movement with the RWM inoperable, TS 3.3.2.1, Control Rod Block Instrumentation, Condition C requires either suspension of control rod movement with the exception of a scram, or one of the following:
x Verification that greater than 12 rods are withdrawn, and verification that movement of control rods is in compliance with banked position withdrawal sequence by a second licensed Operator or other qualified member of the technical staff, OR x
Verification that startup with RWM inoperable has not been performed in the last calendar year, and that movement of control rods is in compliance with banked position withdrawal sequence by a second licensed Operator or other qualified member of the technical staff.
2.3 Reason for the Proposed Change Energy Northwest is requesting approval of an amendment to TS 3.3.2.1, Control Rod Block Instrumentation, Required Action C.2.1.2. Columbia's TS 3.3.2.1 requires an operable RWM in Modes 1 and 2 while less than 10% rated thermal power, and permits only one reactor startup per calendar year with the RWM inoperable, per Condition C. If the RWM is not restored to operable status, TS 3.3.2.1.C must be entered, requiring verification that movement of the control rods is in compliance with the BPWS by a second licensed Operator or other qualified member of the technical staff, and one of the following for continued control rod movement: verification that greater than 12 control rods are withdrawn, or verification that startup with an inoperable RWM has not been performed within the last calendar year. In the event that prolonged RWM issues are experienced, this overly restrictive constraint could result in a challenge in being able to restart the reactor even when the function of the RWM can be fulfilled by the use of a second licensed Operator or qualified member of the technical staff.
The proposed change to Required Action C.2.1.2 is to allow additional reactor startups with a new action to verify FRQWUROrod coupling checks have been performed prior to a reactor restart. This change provides another layer of defense by breaking the sequence of events that would be required to lead to an RDA, rendering this accident as unfeasible. Compliance with regulatory requirements is maintained, and existing margin of safety is maintained.
2.4 Description of the Proposed Change Energy Northwest is proposing to modify Required Action C.2.1.2 from TS 3.3.2.1, which restricts reactor startup with an inoperable RWM to once per calendar year. The proposed method of minimizing the effects of a rod drop accident and maintaining acceptable rod worth will be the existing provisions in TS 3.3.2.1, as well as the utilization of existing station procedures and prescribed control rod pattern templates. The function of the RWM to enforce adherence to the pre-established control rod sequences will continue to be manually
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GO2-25-016 Enclosure Page 3 of 11 fulfilled with the use of a second licensed Operator or a qualified member of the technical staff, as required by Required Action C.2.2 of Condition C.
The proposed changes to TS 3.3.2.1, Condition C, are shown below.
C.
Rod worth minimizer (RWM) inoperable during reactor startup.
C.1 Suspend control rod movement except by scram.
OR C.2.1.1 Verify t 12 rods withdrawn.
OR C.2.1.2 Verify by administrative methods that startup with RWM inoperable has not been performed in the last calendar year.
Verify FRQWUROrod coupling
FKHFNVare performed for first 12 rods.
AND C.2.2 Verify movement of control rods is in compliance with banked position withdrawal sequence (BPWS) by a second licensed operator or other qualified member of the technical staff.
Immediately Immediately Immediately During control rod movement
3.0 TECHNICAL EVALUATION
3.1 Defense-in-Depth Evaluation The control rods inside an operating reactor provide the primary means for control of reactivity changes. To ensure that fuel design limits specified in Columbias Final Safety Analysis Report (FSAR) are not exceeded for postulated transients and accidents, control rod worth is managed by withdrawal sequences and patterns, which are selected prior to operation. The control rod sequences, in conjunction with the velocity limiter, ensure that peak fuel enthalpy remains below the acceptance criteria of 280 cal/g. During high power operation, the rod block monitor provides protection for control rod withdrawal error events.
During low power operations and reactor startup, control rod blocks from the RWM enforce compliance with the pre-established control rod sequences.
The RWM serves as a backup barrier to protect the reactor by preventing Operators from moving an incorrect control rod. It is not intended to replace Operator selection of control patterns but is intended simply to monitor and reinforce procedural adherence. The Operator
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GO2-25-016 Enclosure Page 4 of 11 and pre-established control rod sequences serve as the first-line barrier in preventing the establishment of high worth control rods. Should the RWM become inoperable for any reason, the Operator can maintain acceptable rod worth by adhering to station procedures and prescribed control rod patterns and sequences when below 10% rated thermal power.
In the event the RWM is inoperable, TS 3.3.2.1 Condition C.2.2 requires a second licensed Operator or qualified member of the technical staff to verify that the movement of control rods is in compliance with the BPWS, providing yet another layer of defense-in-depth for precluding a design basis RDA with excessive rod worth such that the 280 cal/g criteria might be exceeded.
The requirement to have an operable RWM for General Electric Boiling Water Reactors (BWR) was introduced during the development of the Improved Standard Technical Specifications. The RWM operability requirements were added to TS to provide a strong incentive for stations to maintain and improve the operability of the RWM when the Rod Sequence Control System requirements were removed from TS. The addition of RWM operability requirements was made to ensure a reasonable degree of operability of the RWM, as at the time this backup system was often poorly maintained and frequently bypassed, providing no significant protection.
It should be noted that the time of the conversion to Improved Standard Technical Specifications precedes 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, which requires that licensees monitor the performance or condition of certain components in a manner to provide reasonable assurance that the intended functions are fulfilled. Although the RWM is a non-safety related system, the ability to bypass the system is relied upon in plant emergency operating procedures, and the all-rod full-in indication provided by the RWM is relied upon to mitigate an Anticipated Transient Without Scram accident, both functions which require monitoring per 10 CFR 50.65(a)(1). In addition to required functional tests, Energy Northwest also evaluates system events under the Maintenance Rule program to further ensure a reasonable degree of operability of the RWM.
In the years since RWM operability requirements were added to TS, continued studies of RDA methodology and results have indicated a substantial reduction in enthalpy for a given rod worth as a result of better core geometry and moderator reactivity modeling. The probability of exceeding the 280 cal/gm fuel enthalpy limit as determined by the Advisory Committee on Reactor Safeguards (Reference 2) could be further reduced by current improved analysis methods and the use of BPWS rod patterns (which reduce the expected maximum rod worth). These results suggest that the RDA probability is likely less than previously analyzed, and that the reliance on the RWM is outdated. The proposed change to TS 3.3.2.1 further reduces the probability of an RDA, as the coupling checks ensure the control rod is properly coupled with its drive mechanism in support of a reactor startup.
3.2 Accident Analysis and Safety Margin In determining the impact of the proposed change against analyzed accidents, Energy Northwest evaluated the RDA and control rod withdrawal error (RWE).
The rapid removal of a high worth control rod in a BWR could result in a potentially significant power excursion. Prevention or mitigation of positive reactivity insertion events, or
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GO2-25-016 Enclosure Page 5 of 11 power excursions, is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage, which could result in undue release of radioactivity. The RDA postulates the de-coupling of a fully inserted control rod from its drive while remaining in the fully inserted position. The scenario then assumes that the dropping of the rod results in a high local reactivity in a small region of the core, and for large, loosely coupled cores, significant shifts in the spatial power generation during the course of the excursion. To limit the worth of the postulated dropped rod and thereby preclude conditions in which dropped rods would have sufficient reactivity to cause the damage assumed by the power excursion, the control rod withdrawal sequences and patterns are selected prior to operation. Operable control rods are required to comply with the requirements of the BPWS when reactor power is less than 10% rated thermal power, to ensure the initial conditions of the RDA analysis are not violated. The RWM does not mitigate or prevent an RDA and is assumed to not function during an RDA. This accident is mitigated by the Average Power Range Monitor, which generates a high flux scram signal to the Reactor Protection System, resulting in an automatic scram of the reactor.
Existing calculations demonstrate that no significant RDA can occur above 10% rated thermal power, with the most severe case occurring when the reactor coolant is at rated temperature with no voids and an out-of-sequence rod is continuously withdrawn. At higher power levels, increased voiding occurs in the core, which flattens the flux profile surrounding a control rod. At this point, the reactivity change from a dropped rod would be significantly less than at the lower power level; therefore, the RDA is not considered at power levels above 10% rated thermal power. The RDA is not expected to occur during the lifetime of the plant and is not modeled in the probabilistic risk assessment for Columbia due to the specific system failures and personnel errors that would have to occur in the correct combination and sequence to present the reactivity required for a design basis RDA to occur. From the NRC Safety Evaluation for Amendment 17 to the General Electric Licensing Topical Report NEDE-24011-P, General Electric Standard Application for Reactor Fuel (Reference 3), the probability for these conditions resulting in a rod drop accident exceeding 280 cal/g peak fuel enthalpy is approximately 10-12 per reactor year. This indicates a large margin to the American Nuclear Society nuclear safety acceptance criterion (Reference 5), which specifies that events with the best estimate frequency of occurrence of 10-6 per reactor year need not be considered for design. The proposed change to require FRQWUROrod coupling checks for the first 12 control rods that are withdrawn upon a reactor startup breaks the chain of events required for an RDA to occur, further decreasing the probability of occurrence.
The inadvertent Operator-initiated withdrawal of a single control rod from the core is classified as a nonlimiting transient event. As discussed in Section 15.4.1.2 of Columbias FSAR, the RWE at low power is categorized as an infrequent accident and is not considered credible during reactor startup or during low power ranges. The probability of development of this event is low because, similar to the RDA, the RWE accident is contingent upon specific failures occurring in a specific sequence; failure of the RWM, and an Operator selecting an out-of-sequence rod with a high worth (contrary to station procedures), as well as disregard of continuous alarm annunciations would have to occur. No mathematical models are involved in this event and consideration of uncertainties is not applicable. In the event the RWM is inoperable during reactor startup, the RWE event is precluded by a second licensed Operator or qualified member of the technical staff verifying that control rod movement complies with the BPWS. For these conditions, the use of a second Operator or
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GO2-25-016 Enclosure Page 6 of 11 qualified member of the technical staff was found to be acceptable in Amendment 149 of Columbias TS (ML17292A737).
Therefore, Energy Northwest has determined that neither the RDA nor the RWE analyses are changed or impacted by the proposed amendment.
4.0 REGULATORY EVALUATION
Chapter 3 of the Columbia FSAR provides detailed discussion of Columbias compliance with the applicable regulatory requirements and guidance.
The proposed TS amendment:
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, independent, redundant, and physically or electrically separated; and x
Does not affect station compliance with the applicable regulations or guidance described in the Columbia FSAR.
4.1 Applicable Regulatory Requirements The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. Energy Northwest has determined that the proposed change does not require any exemptions or relief from regulatory requirements from the following current applicable regulations and regulatory requirements which were reviewed in making this determination.
10 CFR 50.36, Technical Specifications, in which the NRC established its regulatory requirements related to the contents of the TS. Specifically, 10 CFR 50.36(c)(2)(i) states, in part, Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. The proposed change does not affect compliance with these regulations.
10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 13, Instrumentation and Control, states that Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges. The proposed change does not affect compliance with this requirement.
10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 20, Protection System Functions, states that The protection system shall be designed (1) to
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GO2-25-016 Enclosure Page 7 of 11 initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operation occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety. This criterion continues to be met because the proposed change does not impact operation of the reactivity control system, ensuring that fuel design limits are not exceeded.
10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 26, Reactivity Control System Redundancy and Capability, states that Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions. The proposed change does not affect compliance with this requirement because the existing accident analysis remains valid.
10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 28, Reactivity Limits, states that The reactivity control systems shall be designed with appropriate limits on the amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition. The proposed change does not affect compliance with this requirement because the existing accident analysis remains valid.
4.2 Applicable Regulatory Guidance Although the proposed change will not result in any physical modifications to the plant, acceptance criteria and system considerations presented in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, were reviewed as part of developing Section 3.0, Technical Evaluation. Specifically, Standard Review Plan (SRP) 4.3, Nuclear Design (Reference 6), SRP 7.7, Control Systems (Reference 7), and SRP 15.4.9, Spectrum of Rod Drop Accidents (BWR) (Reference 8),
were assessed to ensure the applicable regulatory guidance continues to be met.
5.0 PRECEDENT While there is no approved precedence to permanently eliminate the limitation of one startup per calendar year with an inoperable RWM, the stations listed below received approval to temporarily lift this restriction. The NRC found that the unlimited reactor startups authorized by the amendments could be conducted without endangering the health and safety of the public.
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GO2-25-016 Enclosure Page 8 of 11 x
Oyster Creek Nuclear Generating Station was granted approval on March 11, 1986, for a temporary note allowing unlimited startups with an inoperable RWM during their Operating Cycle 11 (ML011160423). In addition to the use of existing procedural controls, Oyster Creek committed to the use of control rod pattern templates when the RWM was inoperable.
x James A. FitzPatrick Nuclear Power Plant was granted approval on November 19, 2024, for a temporary note that would not restrict reactor startup with an inoperable RWM while compensatory measures were implemented (ML24313A147).
Existing TS requirements for Columbia dictate the use of pre-established sequences for control rods, as well as verification by a second licensed Operator or other qualified member of the technical staff that the rod movement complies with these pre-established sequences when the RWM is inoperable. The proposed change to TS 3.3.2.1 provides yet another layer of defense against a RDA, in that performing the FRQWUROrod coupling checks eliminates the assumption in the RDA analysis that a fully inserted rod becomes uncoupled from its drive.
6.0 NO SIGNIFICANT HAZARDS CONSIDERATION 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment involves no changes to the operation of any system or component in normal or accident operating conditions and does not change the probability or consequences of any station accident.
Station TS will continue to ensure that operable control rods comply with the requirements of the BPWS, and that in the event of an inoperable RWM, an additional licensed Operator or other qualified member of the technical staff enforce compliance with the pre-established rod movement sequence, thereby mitigating any potential unexpected rod withdrawal movement.
The RDA was previously included in the reload analysis; however, it was proposed to the NRC, and subsequently found acceptable, to eliminate the RDA from the standard General Electric Boiling Water Reactor reload package for those plants utilizing the BPWS.
The function of the RWM to perform as a backup to the Operator to inhibit incorrect control rod movement is not required to mitigate any accident conditions. It should be noted that with or without an inoperable RWM, there are no consequences from an RDA that would challenge design limits of the fuel or endanger the health and safety of the public; the radiological consequences resulting from an RDA are bounded by
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GO2-25-016 Enclosure Page 9 of 11 cycle-specific analyses, which have confirmed that the number of fuel rods with an enthalpy greater than the threshold for fuel failure are well below the number assumed in the analysis. The proposed amendment does not impact the calculated exposures from the RDA analysis, which are within the limits of 10 CFR 50.67.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of any accident previously evaluated.
- 2) Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously analyzed?
Response: No.
Additional reactor restarts with an inoperable RWM are not a precursor to a new or different kind of accident, and do not initiate any new or different kinds of accidents.
The required sequence of events for an RDA to occur are not changed by the proposed amendment in a way that would create a new or different kind of accident.
Maintaining control rod worth is principally a manually controlled action, with the Operator following the pre-established sequence. Manual control of rod withdrawals and the established sequences for movement are not being changed with the proposed amendment. The currently approved allowance for independent verification of rod movement by a second licensed Operator or other qualified member of the technical staff to ensure compliance with the BPWS is also not changed with the proposed amendment.
Additionally, the requirement for operable control rods to comply with the BPWS will remain in effect. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously analyzed.
- 3) Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The proposed change does not change the safety function of any plant equipment and does not impact any design basis accident inputs or assumptions. The proposed change does not contribute to, or result in, any changes in which the plant is operated, and does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety.
The margin of safety is established through the design of the plant structures, systems, and components, and administrative controls within which the plant is operated. There are no changes being proposed that decrease the margin of safety.
The margin of safety to the consequences of a control rod drop accident is maintained with the use of additional administrative controls described within the current TS.
Therefore, there is no significant reduction in the margin of safety associated with the proposed amendment.
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GO2-25-016 Enclosure Page 10 of 11 Based on the above, Energy Northwest concludes that the proposed amendment 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 change requirements with respect to installation or use of a facility component located within Columbias restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. Energy Northwest has evaluated the proposed change and has determined that the change does not involve, (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change 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. General Electric Topical Report NEDO-21231, Banked Position Withdrawal Sequence, dated January 1977
- 2. Letter and enclosure from B. C. Rusche (NRR) to R. Fraley (ACRS), Generic Item IIA-2 Control Rod Drop Accident (BWRs), dated June 1, 1976 (ML20125D493, ML20125D495)
- 3. NRC Safety Evaluation, Relating to Amendment 17 General Electric Topical Report NEDE-24011-P, General Electric Standard Application for Reactor Fuel, dated December 27, 1987 (ADAMS Accession Number ML260069M901)
- 4. Letter from T. A. Pickens (BWROG) to NRC, Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, dated August 15, 1986 (ADAMS Accession Number ML20203L984)
- 5. American National Standard Nuclear Safety Criteria for the Design of Stationary Boiling Water Reactor Plants, ANSI/ANS-52.1-1983
- 6. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 4.3, Nuclear Design, Revision 3, March 2007 (ADAMS Accession Number ML070740003)
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GO2-25-016 Enclosure Page 11 of 11
- 7. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 7.7, Control Systems, Revision 6, August 2016 (ADAMS Accession Number ML16020A095)
- 8. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 15.4.9, Spectrum of Rod Drop Accidents (BWR), Revision 3, March 2007 (ADAMS Accession Number ML070880015)
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GO2-25-016 PROPOSED COLUMBIA TECHNICAL SPECIFICATION CHANGE (MARK-UP)
(2 pages follow)
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Control Rod Block Instrumentation 3.3.2.1 Columbia Generating Station 3.3.2.1-1 Amendment No. 169 225 226 253 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 (RBM) channel inoperable.
A.1 Restore RBM channel to OPERABLE status.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> B. Required Action and associated Completion Time of Condition A not met.
OR Two RBM channels inoperable.
B.1 Place one RBM channel in trip.
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. Rod worth minimizer (RWM) inoperable during reactor startup.
C.1 Suspend control rod movement except by scram.
OR C.2.1.1 Verify t 12 rods withdrawn.
OR Immediately Immediately No changes made to this page
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Control Rod Block Instrumentation 3.3.2.1 Columbia Generating Station 3.3.2.1-2 Amendment No. 169 225 226 253 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued)
C.2.1.2 Verify by administrative methods that startup with RWM inoperable has not been performed in the last calendar year.
Verify FRQWUROrod coupling checks are performed for first 12 rods.
AND C.2.2 Verify movement of control rods is in compliance with banked position withdrawal sequence (BPWS) by a second licensed operator or other qualified member of the technical staff.
Immediately During control rod movement D. RWM inoperable during reactor shutdown.
D.1 Verify movement of control rods is in compliance with BPWS by a second licensed operator or other qualified member of the technical staff.
During control rod movement E. One or more Reactor Mode Switch - Shutdown Position channels inoperable.
E.1 Suspend control rod withdrawal.
AND E.2 Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.
Immediately Immediately
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GO2-25-016 REVISED COLUMBIA TECHNICAL SPECIFICATION PAGES (2 pages follow)
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Control Rod Block Instrumentation 3.3.2.1 Columbia Generating Station 3.3.2.1-1 Amendment No. 169 225 226 253 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 (RBM) channel inoperable.
A.1 Restore RBM channel to OPERABLE status.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> B. Required Action and associated Completion Time of Condition A not met.
OR Two RBM channels inoperable.
B.1 Place one RBM channel in trip.
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. Rod worth minimizer (RWM) inoperable during reactor startup.
C.1 Suspend control rod movement except by scram.
OR C.2.1.1 Verify t 12 rods withdrawn.
OR Immediately Immediately No changes made to this page
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Control Rod Block Instrumentation 3.3.2.1 Columbia Generating Station 3.3.2.1-2 Amendment No. 169 225 226 253 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued)
C.2.1.2 Verify FRQWUROrod coupling FKHFNVare performed for first 12 rods.
AND C.2.2 Verify movement of control rods is in compliance with banked position withdrawal sequence (BPWS) by a second licensed operator or other qualified member of the technical staff.
Immediately During control rod movement D. RWM inoperable during reactor shutdown.
D.1 Verify movement of control rods is in compliance with BPWS by a second licensed operator or other qualified member of the technical staff.
During control rod movement E. One or more Reactor Mode Switch - Shutdown Position channels inoperable.
E.1 Suspend control rod withdrawal.
AND E.2 Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.
Immediately Immediately
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GO2-25-016 PROPOSED COLUMBIA TECHNICAL SPECIFICATION BASES CHANGE (MARK-UP FOR INFORMATION ONLY)
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Control Rod Block Instrumentation B 3.3.2.1 Columbia Generating Station B 3.3.2.1-7 Revision 73 BASES ACTIONS A.1 With one RBM channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod block function; however, overall reliability is reduced because a single failure in the remaining OPERABLE channel can result in no control rod block capability for the RBM. For this reason, Required Action A.1 requires restoration of the inoperable channel to OPERABLE status. The Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on the low probability of an event occurring coincident with a failure in the remaining OPERABLE channel.
B.1 If Required Action A.1 is not met and the associated Completion Time has expired, the inoperable channel must be placed in trip within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
If both RBM channels are inoperable, the RBM is not capable of performing its intended function; thus, one channel must also be placed in trip. This initiates a control rod withdrawal block, thereby ensuring that the RBM function is met.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities and is acceptable because it minimizes risk while allowing time for restoration or tripping of inoperable channels.
C.1, C.2.1.1, C.2.1.2, and C.2.2 With the RWM inoperable at any time during a reactor startup, the operator is still capablemaintains capability of enforcing the prescribed control rod sequence. However, the overallOverall reliability is reduced, however, because a single operator error can result in violating the control rod sequence. Therefore, in the event the RWM is inoperable, control rod movement must be immediately suspended except by scram.
Alternatively, sStartup with an inoperable RWM may begin or continue if at least 12 control rods have already been withdrawn, or a reactor startup with an inoperable RWM during withdrawal of one or more of the first 12 rods was not performed in the last calendar yearif FRQWUROrod coupling checks are performed for the first 12 rods. These requirements minimize the number of reactor startups initiated with RWM inoperable. RBoth Required Actions C.2.1.1 and C.2.1.2 require verification of these conditions by review of plant logs andusing control room indications and administrative means. Once Required Action C.2.1.1 or C.2.1.2 is satisfactorily completed, control rod withdrawal may proceed in accordance with the restrictions imposed by Required Action C.2.2.
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