|
|
| Line 18: |
Line 18: |
|
| |
|
| =Text= | | =Text= |
| {{#Wiki_filter:Constellation 4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office 10 CFR 50.90 RS-23-059 June 8, 2023 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265 | | {{#Wiki_filter:}} |
| | |
| ==Subject:==
| |
| License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b"
| |
| | |
| ==Reference:==
| |
| Public meeting between U.S. NRC and Constellation Energy Generation, LLC Regarding Planned License Amendment Request for Quad Cities Nuclear Power Station, Units 1 and 2 on March 2, 2023 (ADAMS Accession Nos. ML23039A145 and ML23068A376)
| |
| In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Constellation Energy Generation, LLC (CEG) requests an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS) Units 1 and 2. The proposed amendment would modify the Technical Specifications (TS) requirements to permit the use of Risk Informed Completion Times (RICTs) in accordance with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times -
| |
| RITSTF Initiative 4b," (ADAMS Accession No. ML18183A493). A model safety evaluation was provided by the NRC to the Technical Specifications Task Force (TSTF) on November 21, 2018 (ADAMS Accession No. ML18253A085 (transmittal letter) and ML18267A259 (model SE)).
| |
| Attachment 1 provides a description and assessment of the proposed changes, the requested confirmation of applicability, and plant-specific verifications. It also provides information supporting the redundancy and diversity of instrumentation governed by the TS proposed to be included as part of the RICT Program in this submittal.
| |
| Attachment 2 provides the existing TS pages marked up to show the proposed changes.
| |
| Attachment 3 provides the existing TS Bases pages marked up to show the proposed changes and is provided for information only.
| |
| Attachment 4 provides a cross-reference between the improved Standard Technical Specifications included in TSTF-505, Rev. 2 and the QCNPS plant-specific TS.
| |
| Attachment 5 provides the RICT program implementation items to be completed prior to program implementation.
| |
| | |
| June 8, 2023 U.S. Nuclear Regulatory Commission Page 2 Attachment 6 provides the proposed License Condition wording for each unit's renewed facility operating license.
| |
| The proposed change has been reviewed and approved by the QCNPS Plant Operations Review Committee in accordance with the requirements of the CEG Quality Assurance Program.
| |
| CEG requests approval of the proposed change by June 7, 2024. The amendment shall be implemented within 180 days of approval. This implementation period will provide adequate time for the affected station documents to be revised using the appropriate change control mechanisms.
| |
| In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), CEG is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.
| |
| There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Ms. Rebecca L. Steinman at (630) 657--2831.
| |
| I declare under penalty of perjury that the foregoing is true and correct. Executed on the 8th day of June 2023.
| |
| Respectfully, Patrick R. Simpson Sr. Manager Licensing Constellation Energy Generation, LLC Attachments:
| |
| : 1. Description and Assessment
| |
| : 2. Proposed Technical Specification Changes (Mark-Ups)
| |
| : 3. Proposed Technical Specification Bases Changes (Mark-Ups)
| |
| (For Information Only)
| |
| : 4. Cross-Reference of TSTF-505 and QCNPS Technical Specifications
| |
| : 5. RICT Program Implementation Items
| |
| : 6. Proposed Renewed Facility Operating License Changes (Mark-ups)
| |
| | |
| ==Enclosures:==
| |
| : 1. List of Revised Required Actions to Corresponding PRA Functions
| |
| : 2. Information Supporting Consistency with Regulatory Guide 1.200, Revision 2
| |
| : 3. Information Supporting Technical Adequacy of PRA Models Without PRA Standards Endorsed by Regulatory Guide 1.200, Revision 2
| |
| | |
| June 8, 2023 U.S. Nuclear Regulatory Commission Page 3
| |
| : 4. Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
| |
| : 5. Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF)
| |
| : 6. Justification of Application of At-Power PRA Models to Shutdown Modes
| |
| : 7. PRA Model Update Process
| |
| : 8. Attributes of the Real-Time Risk Model
| |
| : 9. Key Assumptions and Sources of Uncertainty
| |
| : 10. Program Implementation
| |
| : 11. Monitoring Program
| |
| : 12. Risk Management Action Examples cc: Regional Administrator - NRC Region III NRC Senior Resident Inspector - QCNPS NRC Project Manager, NRR - QCNPS Illinois Emergency Management Agency - Division of Nuclear Safety
| |
| | |
| ATTACHMENT 1 Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Description and Assessment
| |
| | |
| License Amendment Request Attachment 1 Adopt Risk Informed Completion Times TSTF-505 Page 1 of 9 Docket Nos. 50-254 and 50-265 Description and Assessment
| |
| | |
| ==1.0 DESCRIPTION==
| |
| | |
| In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Constellation Energy Generation, LLC (CEG) requests an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS) Units 1 and 2. The proposed amendment would modify the Technical Specifications (TS) requirements to permit the use of Risk Informed Completion Times (RICTs) in accordance with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," (ADAMS Accession No. ML18183A493). A model safety evaluation was provided by the NRC to the Technical Specifications Task Force (TSTF) on November 21, 2018 (ADAMS Accession No. ML18253A085 (transmittal letter) and ML18267A259 (model SE)).
| |
| The proposed amendment would modify the TS requirements related to Completion Times (CTs) for Required Actions to provide the option to calculate a longer, risk-informed CT. A new program, the Risk Informed Completion Time (RICT) Program, is added to TS Section 5.5, Programs and Manuals.
| |
| The methodology for using the RICT Program is described in NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)
| |
| Guidelines," Revision 0, which was approved by the U.S. Nuclear Regulatory Commission (NRC) on May 17, 2007. Adherence to NEI 06-09-A is required by the RICT Program.
| |
| The proposed amendment is consistent with TSTF-505, Revision 2. However, only those Required Actions described in Attachment 4 and Enclosure 1, as reflected in the proposed TS markups provided in Attachment 2, are proposed to be changed. Some of the modified Required Actions in TSTF-505 are not applicable to QCNPS. There are also some plant-specific Required Actions not included in TSTF-505 that are included in this proposed amendment.
| |
| 2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation CEG has reviewed TSTF-505, Revision 2, and the model safety evaluation dated November 21, 2018 (ADAMS Accession No. ML18253A085). This review included the information provided to support TSTF-505 and the safety evaluation for NEI 06-09-A. As described in the subsequent paragraphs, CEG has concluded that the technical basis is applicable to QCNPS Units 1 and 2, and support incorporation of this amendment in the QCNPS TS.
| |
| 2.2 Verifications and Regulatory Commitments In accordance with Section 4.0, Limitations and Conditions, of the safety evaluation for NEI 06-09-A, the following is provided:
| |
| : 1. Enclosure 1 identifies each of the TS Required Actions to which the RICT Program will apply, with a comparison of the TS functions to the functions modeled in the probabilistic risk assessment (PRA) of the structures, systems and components (SSCs) subject to those actions.
| |
| | |
| License Amendment Request Attachment 1 Adopt Risk Informed Completion Times TSTF-505 Page 2 of 9 Docket Nos. 50-254 and 50-265 Description and Assessment
| |
| : 2. Enclosure 2 provides a discussion of the results of peer reviews and self-assessments conducted for the plant-specific PRA models which support the RICT Program, as specified in Regulatory Guide (RG) 1.200, Section 4.2.
| |
| : 3. Enclosure 3 is not applicable since each PRA model used for the RICT Program is addressed using a standard endorsed by the NRC. However, a cover page is included for this Enclosure to maintain numbering consistency with the Traveler.
| |
| : 4. Enclosure 4 provides appropriate justification for excluding sources of risk not addressed by the PRA models.
| |
| : 5. Enclosure 5 provides the plant-specific baseline core damage frequency (CDF) and large early release frequency (LERF) to confirm that the potential risk increases allowed under the RICT Program are acceptable.
| |
| : 6. Enclosure 6 is not applicable since the RICT Program is not being applied to shutdown modes. However, a cover page is included for this Enclosure to maintain numbering consistency with the Traveler.
| |
| : 7. Enclosure 7 provides a discussion of CEGs programs and procedures that assure the PRA models that support the RICT Program are maintained consistent with the as-built, as-operated plant.
| |
| : 8. Enclosure 8 provides a description of how the baseline PRA model, which calculates average annual risk, is evaluated and modified for use in the Real-Time Risk (RTR) tool to assess real-time configuration risk, and describes the scope of, and quality controls applied to, the RTR tool.
| |
| : 9. Enclosure 9 provides a discussion of how the key assumptions and sources of uncertainty in the PRA models were identified, and how their impact on the RICT Program was assessed and dispositioned.
| |
| : 10. Enclosure 10 provides a description of the implementing programs and procedures regarding the plant staff responsibilities for the RICT Program implementation, including risk management action (RMA) implementation.
| |
| : 11. Enclosure 11 provides a description of the implementation and monitoring program as described in NEI 06-09-A, Section 2.3.2, Step 7.
| |
| : 12. Enclosure 12 provides a description of the process to identify and provide RMAs.
| |
| 2.3 QCNPS Design Features of Note QCNPS Units 1 and 2 are designed such that each unit operates independently of the other such that a malfunction of equipment or operator error in either of the two units will not affect the continued operation of the other unit. In those instances where a system serving one unit is interconnected with its counterpart in the other unit, the effect of the intertie on the function of
| |
| | |
| License Amendment Request Attachment 1 Adopt Risk Informed Completion Times TSTF-505 Page 3 of 9 Docket Nos. 50-254 and 50-265 Description and Assessment each system is evaluated. Interconnected auxiliary systems are described in the Updated Final Safety Analysis Report (UFSAR) Section 1.2.4.4, with the following common systems that are shared between the two units: the standby gas treatment system (SGTS), the turbine building, portable makeup demineralizer system, radioactive waste system, the safe shutdown makeup pump (SSMP), and other facilities that are not critical to the safe startup, operation, or shutdown on the plant. In addition to these common/shared systems, QCNPS has some design features that are unique compared to the typical boiling water reactor (BWR) design for which NUREG-1433 is based. The following sections briefly summarize the unique features that impact the application of the TSTF-505 Traveler at QCNPS.
| |
| 2.3.1 Safe Shutdown Makeup Pump The SSMP system was installed as a backup to the reactor core isolation cooling (RCIC) system to meet 10 CFR 50, Appendix R, Section III.G requirements. As described in UFSAR Section 5.4.6.5, the SSMP system provides cooling water from the contaminated condensate storage tank (CCST) to the Unit 1 or Unit 2 reactor core if the reactor becomes isolated from the main condenser simultaneously with a loss of feedwater system. The pump discharge header is divided into two segments which inject into the feedwater piping of Unit 1 and the high pressure coolant injection (HPCI) piping of Unit 2. The SSMP system injection valves are interlocked to allow injection into only one reactor vessel at a time. The backup water supply for the SSMP is the fire header, which may be aligned in response to low CCST water level.
| |
| 2.3.2 Standby Auxiliary Source of Electrical Power The QCNPS onsite AC power system consists of two main generators, two main step-up transformers, two unit auxiliary transformers (UATs), two reserve auxiliary transformers (RATs),
| |
| distribution buses, three standby emergency diesel generators (EDGs), and two standby station blackout (SBO) diesel generators (DGs). Enclosure 1 Section 7 provides a more detailed explanation of the electrical system capabilities.
| |
| In response to SBO Rule (i.e., 10 CFR 50.63), QCNPS installed an alternative AC system (AAC). The AAC is a non-class 1E, independent source of additional on-site emergency AC power system consisting of two manually started and manually connected diesel-driven generator sets. The SBO DGs and auxiliaries are located in the Station Blackout Building to protect the system against weather-related events which could initiate an SBO event and provides physical isolation from safety-related components. The building is physically separated from emergency systems, thus avoiding the consequences of multiple failures of on-site diesel generator systems due to severe weather-related events. Each DG is physically separated from the other within the building. Additional details of this system are provided in UFSAR Section 8.3.1.9.
| |
| 2.3.3 FLEX Equipment Modeled in the PRA QCNPS also has FLEX equipment to provide independent means capable of mitigating a simultaneous loss of all AC power and loss of normal access to the ultimate heat sink resulting from a beyond design basis external event (BDBEE) by providing adequate capability to maintain or restore core cooling, containment cooling, and spent fuel pool (SFP) cooling
| |
| | |
| License Amendment Request Attachment 1 Adopt Risk Informed Completion Times TSTF-505 Page 4 of 9 Docket Nos. 50-254 and 50-265 Description and Assessment capabilities to both units. Enclosure 9 Section 2 provides additional details on the use of FLEX equipment in the PRA Models.
| |
| 2.4 Optional Changes and Variations CEG is proposing the following variations from the TS changes described in TSTF-505, Revision 2, or the applicable parts of the NRC model safety evaluation dated November 21, 2018. These options were recognized as acceptable variations in TSTF-505 and the NRC model safety evaluation.
| |
| In a few instances, the QCNPS TS use different numbering and/or titles than the Standard Technical Specifications (STS) on which TSTF-505 was based. These differences are administrative and do not affect the applicability of TSTF-505 to the QCNPS TS. Attachment 4 is a cross-reference that provides a comparison between the NUREG-1433, "Standard Technical Specifications General Electric BWR/4 Plants," Required Actions included in TSTF-505 and the QCNPS Required Actions included in this license amendment request. The attachment includes a summary description of the referenced Required Actions, which is provided for information purposes only and is not intended to be a verbatim description of the Required Action. The cross-reference in Attachment 4 identifies the following:
| |
| : 1. QCNPS Required Actions that have identical numbers to the corresponding NUREG-1433 Required Actions are not deviations from TSTF-505, except for administrative deviations (if any) such as formatting. These deviations are administrative with no impact on the NRC's model safety evaluation dated November 21, 2018.
| |
| : 2. QCNPS Required Actions that have different numbering than the NUREG-1433 Required Actions are an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
| |
| : 3. The QCNPS TS conversion to the improved TS retained elements of the original TS that were consistent with the QCNPS licensing basis and differ from NUREG-1433. For NUREG-1433 Required Actions that are not contained in the QCNPS TS, the corresponding TSTF-505 mark-ups for the Required Actions are not applicable to QCNPS. This is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
| |
| : 4. The model application provided in TSTF-505, Revision 2, includes an attachment for typed, camera-ready (revised) TS pages reflecting the proposed changes. QCNPS is not including such an attachment due to the number of TS pages included in this submittal that have the potential to be affected by other unrelated license amendment requests and the straightforward nature of the proposed changes. Providing only mark-ups of the proposed TS changes satisfies the requirements of 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," in that the mark-ups fully describe the changes desired. This is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018. Because of this deviation, the contents and numbering of the
| |
| | |
| License Amendment Request Attachment 1 Adopt Risk Informed Completion Times TSTF-505 Page 5 of 9 Docket Nos. 50-254 and 50-265 Description and Assessment attachments for this amendment request differ from the attachments specified in the model application in TSTF-505.
| |
| : 5. For several of the Instrumentation Section 3.3 TS, as noted in the Attachment 2 TS markups, a footnote is added to the proposed statement "or in accordance with the Risk Informed Completion Time Program" to ensure that a RICT is not applied when the actuation/trip function is lost. Under this circumstance, TSTF-505, Revision 2, specifies the addition of a Note that reads "Not applicable when [all] required [channels] are inoperable." Because the loss of function is dependent upon not only the number of inoperable channels, but also the combination of inoperable channels within the trip systems, CEG has chosen to replace the TSTF-505 Note with a footnote which reads "Not applicable when trip capability is not maintained," which accomplishes the intended purpose of the TSTF-505 Note.
| |
| : 6. There are two plant-specific design specific LCOs and associated Required Actions for which QCNPS is proposing to apply the RICT Program that are variations from TSTF-505, Revision 2, as identified in Attachment 4. Attachment 4 was created using the BWR/4 standard from NUREG-1433, with exceptions annotated in Attachment 4 and summarized below. Additional details are contained in Attachment 4 for TS Conditions and Required Actions.
| |
| TS 3.6.2.6 - Residual Heat Removal (RHR) Drywell Spray - QCNPS TS 3.6.2.6 Condition A: One drywell spray subsystem inoperable.
| |
| Drywell spray is a mode of the RHR system which may be initiated under post-loss-of-coolant accident (LOCA) conditions to remove inorganic iodines and particulates from the drywell atmosphere by washing, or scrubbing, them into the suppression pool. This function, which is credited in the LOCA radiological dose analysis, reduces the amount of airborne activity available for leakage from the drywell to ensure that the radiological consequences from the accident remain within the limits of 10 CFR 50.67. This function is provided by two redundant drywell spray subsystems. With one RHR drywell spray subsystem inoperable, the remaining operable RHR drywell spray subsystem is adequate to perform the primary containment fission product scrubbing function. Application of a RICT is acceptable considering the redundant RHR drywell spray capabilities afforded by the operable subsystem and the low probability of a LOCA occurring during this period.
| |
| TS 3.7.9 - Safe Shutdown Makeup Pump (SSMP) System - QCNPS TS 3.7.9 Condition A is a plant-specific Condition not in the NUREG-1433 TS, and therefore, not in TSTF-505, Revision 2.
| |
| The QCNPS SSMP is a common unit system designed to supply makeup water to the reactor core at the same capacity as the RCIC system. The SSMP System design requirements ensure that the criteria of 10 CFR 50, Appendix R, Section III.G are satisfied. See Section 2.3.1 of this attachment for additional details regarding the SSMP system.
| |
| | |
| License Amendment Request Attachment 1 Adopt Risk Informed Completion Times TSTF-505 Page 6 of 9 Docket Nos. 50-254 and 50-265 Description and Assessment Condition A applies when the system is inoperable. Required Action A.1 allows 14 days to restore the system to an operable status. In this Condition, loss of the SSMP System will not affect the overall plant capability to provide makeup inventory at high reactor pressure since the RCIC and HPCI System are required to be operable.
| |
| As indicated in Enclosure 1, the SSMP system is explicitly modeled in the QCNPS PRA. The PRA Success Criterion is that one of one train can provide RPV makeup during RPV isolation with loss of feedwater flow.
| |
| Therefore, TS 3.7.9 Condition A meet the requirements for inclusion in the RICT Program.
| |
| CEG has determined that the application of a RICT for these QCNPS plant-specific LCOs is consistent with TSTF-505, Revision 2, and with the NRC's model safety evaluation dated November 21, 2018. Application of a RICT for these plant-specific LCOs will be controlled under the RICT Program. The RICT Program provides the necessary administrative controls to permit extension of Completion Times and thereby delay reactor shutdown or remedial actions if risk is assessed and managed within specified limits and programmatic requirements. The specified safety function or performance levels of TS required SSCs are unchanged, and the remedial actions, including the requirement to shutdown the reactor, are also unchanged; only the Completion Times are extended by the RICT Program.
| |
| Application of a RICT will be evaluated using the methodology and probabilistic risk guidelines contained in NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," which was approved by the NRC on May 17, 2007 (ADAMS Accession No. ML071200238). The NEI 06-09-A, methodology includes a requirement to perform a quantitative assessment of the potential impact of the application of a RICT on risk, to reassess risk due to plant configuration changes, and to implement compensatory measures and risk management actions (RMAs) to maintain the risk below acceptable regulatory risk thresholds. In addition, the NEI 06-09-A, methodology satisfies the five key safety principles specified in Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decision making: Technical Specifications," Revision 2, dated January 2021 (ADAMS Accession No. ML20164A034), relative to the risk impact due to the application of a RICT.
| |
| Therefore, the proposed application of a RICT in the QCNPS plant-specific Required Actions is consistent with TSTF-505, Revision 2, and with the NRC model safety evaluation dated November 21, 2018.
| |
| CEG has reviewed these changes and determined that they do not affect the applicability of TSTF-505, Revision 2, to the QCNPS TS.
| |
| | |
| License Amendment Request Attachment 1 Adopt Risk Informed Completion Times TSTF-505 Page 7 of 9 Docket Nos. 50-254 and 50-265 Description and Assessment Additionally, QCNPS was not licensed to the 10 CFR 50, Appendix A, General Design Criteria (GDC). QCNPS UFSAR Section 3.1 "Conformance with NRC Design Criteria," provides an assessment against the 70 draft GDC published in 1967 and concludes that the plant-specific requirements are sufficiently similar to the Appendix A GDC. Therefore, this difference does not alter the conclusion TSTF-505, Revision 2 is applicable to QCNPS.
| |
| | |
| ==3.0 REGULATORY ANALYSIS==
| |
| | |
| 3.1 No Significant Hazards Consideration Determination Constellation Energy Generation, LLC (CEG) has evaluated the proposed changes to the Technical Specifications (TS) using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.
| |
| Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, request adoption of an approved change to the Standard Technical Specifications (STS) and plant-specific TS, to modify the TS requirements related to Completion Times for Required Actions to provide the option to calculate a longer, risk-informed Completion Time. The allowance is described in a new program in Chapter 5.0, "Administrative Controls," entitled the "Risk Informed Completion Time Program."
| |
| As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:
| |
| : 1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?
| |
| Response: No.
| |
| The proposed changes permit the extension of Completion Times provided the associated risk is assessed and managed in accordance with the NRC approved Risk Informed Completion Time Program. The proposed changes do not involve a significant increase in the probability of an accident previously evaluated because the changes involve no change to the plant or its modes of operation. The proposed changes do not increase the consequences of an accident because the design-basis mitigation function of the affected systems is not changed and the consequences of an accident during the extended Completion Time are no different from those during the existing Completion Time.
| |
| Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
| |
| : 2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
| |
| Response: No.
| |
| | |
| License Amendment Request Attachment 1 Adopt Risk Informed Completion Times TSTF-505 Page 8 of 9 Docket Nos. 50-254 and 50-265 Description and Assessment The proposed change does not change the design, configuration, or method of operation of the plant, except with respect to the flood mitigation strategy.
| |
| The current flood mitigation strategy allows the flood waters to enter the plant as soon as the level reaches a ground floor elevation of 595. The station will delay water entry into the plant by installing local intense precipitation (LIP) barriers. With minor modifications, the LIP barriers will be capable of providing protection up to a water surface elevation of 599'. Additional verifications were performed to ensure that all credited flood boundary elements (e.g., building walls, exterior doors, etc.) can withstand the hydrostatic and hydrodynamic forces from the additional water accumulation at the higher flood level.
| |
| The proposed modification only delays the time of flood water entry into the station without altering the mitigation strategy. The functionality, design, and implementation of the LIP barriers remains unchanged, except for minor upgrades such as increasing the height of one barrier. The current Updated Final Safety Analysis Report (UFSAR) described flood mitigation strategy remains available to mitigate the higher magnitude, low frequency floods that exceed 599' and overtop the barriers.
| |
| Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
| |
| : 3. Do the proposed changes involve a significant reduction in a margin of safety?
| |
| Response: No.
| |
| The proposed changes permit the extension of Completion Times provided that risk is assessed and managed in accordance with the NRC approved Risk Informed Completion Time Program. The proposed changes implement a risk-informed configuration management program to assure that adequate margins of safety are maintained. Application of these new specifications and the configuration management program considers cumulative effects of multiple systems or components being out of service and does so more effectively than the current TS.
| |
| Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
| |
| Based on the above, CEG concludes that the proposed changes present no 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.
| |
| 3.2 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
| |
| | |
| License Amendment Request Attachment 1 Adopt Risk Informed Completion Times TSTF-505 Page 9 of 9 Docket Nos. 50-254 and 50-265 Description and Assessment 4.0 ENVIRONMENTAL EVALUATION The proposed changes would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed changes.
| |
| | |
| ATTACHMENT 2 Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Proposed Technical Specification Changes (Mark-Ups)
| |
| TS Pages 3.6.1.2-3 1.3-13 3.6.1.3-1 and -2 3.1.7-1 3.6.1.3-4 3.3.1.1-1 3.6.1.6-1 3.3.2.2-1 3.6.1.7-1 and -2 3.3.4.1-1 3.6.1.8-1 3.3.5.1-2 thru -7 3.6.2.3-1 3.3.5.3-1 and -2 3.6.2.6-1 3.3.6.1-1 3.7.1-1 3.3.6.3-1 3.7.9-1 3.3.8.1-1 3.8.1-2 thru -4 3.4.3-1 3.8.4-1 thru -3 3.5.1-1 and -2 3.8.7-1 and -2 3.5.3-1 5.5-14
| |
| | |
| Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued) is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expired.
| |
| Insert A here IMMEDIATE When "Immediately" is used as a Completion Time, the COMPLETION TIME Required Action should be pursued without delay and in a controlled manner.
| |
| Quad Cities 1 and 2 1.3-13 Amendment No. 199/195
| |
| | |
| SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE. delete APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable. to OPERABLE status.
| |
| Insert B here B. Two SLC subsystems B.1 Restore one SLC 8 hours inoperable. subsystem to OPERABLE status.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours delete associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours Quad Cities 1 and 2 3.1.7-1 Amendment No. 233/229
| |
| | |
| 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 delete
| |
| ------------------------------------ NOTES------------------------------------
| |
| : 1. Separate Condition entry is allowed for each channel.
| |
| : 2. When Functions 2.b and 2.c channels are inoperable due to the calculated power exceeding the APRM output by more than 2% RTP while operating at 25% RTP, entry into associated Conditions and Required Actions may be delayed for up to 2 hours.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in 12 hours channels inoperable. trip.
| |
| Insert C here OR A.2 Place associated trip 12 hours system in trip.
| |
| Insert C here B. One or more Functions B.1 Place channel in one 6 hours with one or more trip system in trip.
| |
| required channels Insert C here inoperable in both OR trip systems.
| |
| B.2 Place one trip system 6 hours in trip.
| |
| Insert C here (continued)
| |
| Quad Cities 1 and 2 3.3.1.1-1 Amendment No. 272/267
| |
| | |
| Feedwater System and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 3.3 INSTRUMENTATION 3.3.2.2 Feedwater System and Main Turbine High Water Level Trip Instrumentation LCO 3.3.2.2 Four channels of Feedwater System and main turbine high water level trip instrumentation shall be OPERABLE.
| |
| delete APPLICABILITY: THERMAL POWER 25% RTP.
| |
| -----------------------------------NOTE---------------------------------------
| |
| Separate Condition entry is allowed for each channel.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Feedwater A.1 ---------NOTE-------- delete System and main Not applicable if turbine high water inoperable channel is level trip channels the result of an inoperable. inoperable feedwater pump breaker.
| |
| Place channel in 7 days Trip.
| |
| Insert C here B. Feedwater System and B.1 Restore trip 2 hours main turbine high capability.
| |
| water level trip capability not maintained.
| |
| (continued)
| |
| Quad Cities 1 and 2 3.3.2.2-1 Amendment No. 230/225
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.1 3.3 INSTRUMENTATION 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation LCO 3.3.4.1 Two channels per trip system for each ATWS-RPT instrumentation Function listed below shall be OPERABLE:
| |
| : a. Reactor Vessel Water LevelLow Low; and
| |
| : b. Reactor Vessel Steam Dome PressureHigh.
| |
| APPLICABILITY: MODE 1.
| |
| ACTIONS
| |
| -------------------------------------NOTE-------------------------------------
| |
| Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Restore channel to 14 days inoperable. OPERABLE status.
| |
| Insert C here OR A.2 --------NOTE---------
| |
| Not applicable if inoperable channel is the result of an inoperable breaker.
| |
| Place channel in 14 days trip.
| |
| Insert C here (continued)
| |
| Quad Cities 1 and 2 3.3.4.1-1 Amendment No. 199/195
| |
| | |
| ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME delete B. As required by B.1 --------NOTE---------
| |
| Required Action A.1 Only applicable for and referenced in Functions 1.a, 1.b, Table 3.3.5.1-1. 2.a, 2.b, 2.d, and 2.j.
| |
| Declare supported 1 hour from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature(s) in both divisions AND B.2 --------NOTE---------
| |
| Only applicable for Functions 3.a and 3.b.
| |
| Declare High Pressure 1 hour from Coolant Injection discovery of (HPCI) System loss of HPCI inoperable. initiation capability AND B.3 Place channel in 24 hours trip.
| |
| Insert C here (continued)
| |
| Quad Cities 1 and 2 3.3.5.1-2 Amendment No. 273/268
| |
| | |
| ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. As required by C.1 --------NOTE--------- delete Required Action A.1 Only applicable for and referenced in Functions 1.c, 1.e, Table 3.3.5.1-1. 2.c, 2.e, 2.g, 2.h, 2.i, and 2.k.
| |
| Declare supported 1 hour from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. feature(s) in both divisions AND C.2 Restore channel to 24 hours OPERABLE status.
| |
| Insert C here (continued)
| |
| Quad Cities 1 and 2 3.3.5.1-3 Amendment No. 273/268
| |
| | |
| ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. As required by D.1 --------NOTE---------
| |
| Required Action A.1 Only applicable if and referenced in HPCI pump suction is Table 3.3.5.1-1. not aligned to the suppression pool.
| |
| Declare HPCI System 1 hour from inoperable. discovery of loss of HPCI initiation capability AND D.2.1 Place channel in 24 hours trip.
| |
| Insert C here OR D.2.2 Align the HPCI pump 24 hours suction to the suppression pool.
| |
| (continued)
| |
| Quad Cities 1 and 2 3.3.5.1-4 Amendment No. 199/195
| |
| | |
| ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. As required by E.1 --------NOTE---------
| |
| delete Required Action A.1 Only applicable for and referenced in Functions 1.d Table 3.3.5.1-1. and 2.f.
| |
| Declare supported 1 hour from feature(s) inoperable discovery of when its redundant loss of feature ECCS initiation initiation capability capability for is inoperable. subsystems in both divisions AND E.2 Restore channel to 7 days OPERABLE status.
| |
| Insert C here (continued)
| |
| Quad Cities 1 and 2 3.3.5.1-5 Amendment No. 273/268
| |
| | |
| ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. As required by F.1 Declare Automatic 1 hour from Required Action A.1 Depressurization discovery of and referenced in System (ADS) valves loss of ADS Table 3.3.5.1-1. inoperable. initiation capability in both trip systems AND F.2 Place channel in 96 hours from trip. discovery of inoperable channel concurrent with HPCI or reactor core isolation cooling (RCIC) inoperable Insert C here AND 8 days (continued)
| |
| Insert C here Quad Cities 1 and 2 3.3.5.1-6 Amendment No. 199/195
| |
| | |
| ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G. As required by G.1 Declare ADS valves 1 hour from Required Action A.1 inoperable. discovery of and referenced in loss of ADS Table 3.3.5.1-1. initiation capability in both trip systems AND G.2 Restore channel to 96 hours from OPERABLE status. discovery of inoperable channel concurrent with HPCI or RCIC inoperable Insert C here AND 8 days Insert C here H. Required Action and H.1 Declare associated Immediately associated Completion supported feature(s)
| |
| Time of Condition B, inoperable.
| |
| C, D, E, F, or G not met.
| |
| Quad Cities 1 and 2 3.3.5.1-7 Amendment No. 199/195
| |
| | |
| RCIC System Instrumentation 3.3.5.3 3.3 INSTRUMENTATION 3.3.5.3 Reactor Core Isolation Cooling (RCIC) System Instrumentation LCO 3.3.5.3 The RCIC System instrumentation for each Function in Table 3.3.5.3-1 shall be OPERABLE.
| |
| delete APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig.
| |
| ACTIONS
| |
| -------------------------------------NOTE-------------------------------------
| |
| Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Enter the Condition Immediately channels inoperable. referenced in Table 3.3.5.3-1 for the channel.
| |
| B. As required by B.1 Declare RCIC System 1 hour from Required Action A.1 inoperable. discovery of and referenced in loss of RCIC Table 3.3.5.3-1. initiation capability AND B.2 Place channel in 24 hours trip.
| |
| Insert C here (continued)
| |
| Quad Cities 1 and 2 3.3.5.3-1 Amendment No. 273/268
| |
| | |
| RCIC System Instrumentation 3.3.5.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. As required by C.1 Restore channel to 24 hours delete Required Action A.1 OPERABLE status.
| |
| and referenced in Table 3.3.5.3-1.
| |
| D. As required by D.1 --------NOTE---------
| |
| Required Action A.1 Only applicable if and referenced in RCIC pump suction is Table 3.3.5.3-1. not aligned to the suppression pool.
| |
| Declare RCIC System 1 hour from inoperable. discovery of loss of RCIC initiation capability AND D.2.1 Place channel in 24 hours trip.
| |
| Insert C here OR D.2.2 Align RCIC pump 24 hours suction to the suppression pool.
| |
| E. Required Action and E.1 Declare RCIC System Immediately associated Completion inoperable.
| |
| Time of Condition B, C, or D not met.
| |
| Quad Cities 1 and 2 3.3.5.3-2 Amendment No. 273/268
| |
| | |
| Primary Containment Isolation Instrumentation 3.3.6.1 3.3 INSTRUMENTATION 3.3.6.1 Primary Containment Isolation Instrumentation LCO 3.3.6.1 The primary containment isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERABLE.
| |
| APPLICABILITY: According to Table 3.3.6.1-1.
| |
| 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 12 hours for channels inoperable. trip. Functions 1.a, 2.a, 2.b, 3.d, 5.b, and 6.b AND Insert C here 24 hours for Functions other than Functions 1.a, 2.a, 2.b, 3.d, 5.b, and 6.b Insert C here B. One or more automatic B.1 Restore isolation 1 hour Functions with capability.
| |
| isolation capability not maintained.
| |
| (continued)
| |
| Quad Cities 1 and 2 3.3.6.1-1 Amendment No. 199/195
| |
| | |
| Relief Valve Instrumentation 3.3.6.3 3.3 INSTRUMENTATION 3.3.6.3 Relief Valve Instrumentation LCO 3.3.6.3 The relief valve instrumentation for each Function in Table 3.3.6.3-1 shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One relief valve A.1 Restore channel(s) to 14 days inoperable due to OPERABLE status.
| |
| inoperable channel(s). Insert C here B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met.
| |
| B.2 Be in MODE 4. 36 hours OR Two or more relief valves inoperable due to inoperable channels.
| |
| Quad Cities 1 and 2 3.3.6.3-1 Amendment No. 199/195
| |
| | |
| LOP Instrumentation 3.3.8.1 3.3 INSTRUMENTATION 3.3.8.1 Loss of Power (LOP) Instrumentation LCO 3.3.8.1 The LOP instrumentation for each Function in Table 3.3.8.1-1 shall be OPERABLE.
| |
| delete APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS
| |
| -------------------------------------NOTE-------------------------------------
| |
| Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in 1 hour inoperable. trip.
| |
| Insert C here B. Required Action and B.1 Declare associated Immediately associated Completion diesel generator (DG)
| |
| Time not met. inoperable.
| |
| Quad Cities 1 and 2 3.3.8.1-1 Amendment No. 287/283
| |
| | |
| Safety and Relief Valves 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety and Relief Valves LCO 3.4.3 The safety function of 9 safety valves shall be OPERABLE.
| |
| AND The relief function of 5 relief valves shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One relief valve A.1 Restore the relief 14 days inoperable. valve to OPERABLE status. Insert B here B. Required Action and --------------NOTE-----------
| |
| associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.
| |
| not met. ----------------------------- delete B.1 Be in MODE 3. 12 hours C. Two or more relief C.1 Be in MODE 3. 12 hours valves inoperable.
| |
| AND OR C.2 Be in MODE 4. 36 hours One or more safety valves inoperable.
| |
| Quad Cities 1 and 2 3.4.3-1 Amendment No. 245/240
| |
| | |
| ECCSOperating 3.5.1 delete 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)
| |
| WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC)
| |
| SYSTEM 3.5.1 ECCSOperating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of five relief valves shall be OPERABLE.
| |
| APPLICABILITY: MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and 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 HPCI.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One Low Pressure A.1 Restore LPCI pump to 30 days Coolant Injection OPERABLE status.
| |
| (LPCI) pump inoperable.
| |
| B. One LPCI subsystem B.1 Restore low pressure 7 days inoperable for reasons ECCS injection/spray other than Condition subsystem to OPERABLE Insert B here A. status.
| |
| OR One Core Spray subsystem inoperable.
| |
| C. One LPCI pump in each C.1 Restore one LPCI pump 7 days subsystem inoperable. to OPERABLE status.
| |
| Insert B here D. Required Action and --------------NOTE------------
| |
| associated Completion LCO 3.0.4.a is not applicable Time of Condition A, when entering MODE 3.
| |
| B, or C not met. ------------------------------
| |
| D.1 Be in MODE 3. 12 hours (continued)
| |
| Quad Cities 1 and 2 3.5.1-1 Amendment No. 273/268
| |
| | |
| ECCSOperating 3.5.1 delete ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. Two LPCI subsystems E.1 Restore one LPCI 72 hours inoperable for reasons subsystem to OPERABLE other than Condition status. Insert B here C.
| |
| F. Required Action and F.1 Be in MODE 3. 12 hours associated Completion Time of Condition E AND not met.
| |
| F.2 Be in MODE 4. 36 hours G. HPCI System G.1 Verify by Immediately inoperable. administrative means RCIC System is OPERABLE.
| |
| AND G.2 Restore HPCI System 14 days to OPERABLE status.
| |
| Insert B here H. One ADS valve H.1 Restore ADS valve to 14 days inoperable. OPERABLE status.
| |
| Insert B here I. Required Action and --------------NOTE-----------
| |
| associated Completion LCO 3.0.4.a is not applicable Time of Condition G or when entering MODE 3.
| |
| H not met. -----------------------------
| |
| I.1 Be in MODE 3. 12 hours J. Two or more ADS valves J.1 Be in MODE 3. 12 hours inoperable.
| |
| AND J.2 Reduce reactor steam 36 hours dome pressure to
| |
| < 150 psig.
| |
| (continued)
| |
| Quad Cities 1 and 2 3.5.1-2 Amendment No. 245/240
| |
| | |
| RCIC System 3.5.3 delete 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)
| |
| WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC)
| |
| SYSTEM 3.5.3 RCIC System LCO 3.5.3 The RCIC System shall be OPERABLE.
| |
| APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig.
| |
| ACTIONS
| |
| ----------------------------------NOTE---------------------------------
| |
| LCO 3.0.4.b is not applicable to RCIC.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System A.1 Verify by Immediately inoperable. administrative means High Pressure Coolant Injection System is OPERABLE.
| |
| AND A.2 Restore RCIC System 14 days to OPERABLE status.
| |
| Insert B here B. Required Action and --------------NOTE-----------
| |
| associated Completion LCO 3.0.4.a is not applicable Time not met. when entering MODE 3.
| |
| B.1 Be in MODE 3. 12 hours Quad Cities 1 and 2 3.5.3-1 Amendment No. 273/268
| |
| | |
| Primary Containment Air Lock 3.6.1.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Lock an OPERABLE door 24 hours closed.
| |
| AND B.3 --------NOTE---------
| |
| Air lock doors in high radiation areas or areas with limited access due to inerting may be verified locked closed by administrative means.
| |
| Verify an OPERABLE Once per 31 days door is locked closed.
| |
| C. Primary containment C.1 Initiate action to Immediately air lock inoperable evaluate primary for reasons other than containment overall Condition A or B. leakage rate per LCO 3.6.1.1, using current air lock test results.
| |
| AND C.2 Verify a door is 1 hour closed.
| |
| AND C.3 Restore air lock to 24 hours OPERABLE status.
| |
| Insert D here (continued)
| |
| Quad Cities 1 and 2 3.6.1.2-3 Amendment No. 199/195
| |
| | |
| PCIVs 3.6.1.3 3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)
| |
| LCO 3.6.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE. delete APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS
| |
| ------------------------------------- NOTES -----------------------------------
| |
| : 1. Penetration flow paths may be unisolated intermittently under administrative controls.
| |
| : 2. Separate Condition entry is allowed for each penetration flow path.
| |
| : 3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.
| |
| : 4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment," when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. ---------NOTE-------- A.1 Isolate the affected 4 hours except Only applicable to penetration flow path for main steam penetration flow paths by use of at least line with two or more one closed and PCIVs. de-activated AND Insert B here
| |
| --------------------- automatic valve, closed manual valve, 8 hours for main One or more blind flange, or steam line penetration flow paths check valve with flow with one PCIV through the valve Insert B here inoperable for reasons secured.
| |
| other than Condition D. AND (continued)
| |
| Quad Cities 1 and 2 3.6.1.3-1 Amendment No. 287/283
| |
| | |
| PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 --------NOTES--------
| |
| : 1. Isolation devices in high radiation areas may be verified by use of administrative means.
| |
| : 2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.
| |
| --------------------- following isolation Verify the affected Once per 31 days penetration flow path for isolation is isolated. devices outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment (continued)
| |
| Quad Cities 1 and 2 3.6.1.3-2 Amendment No. 199/195
| |
| | |
| PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 --------NOTES--------
| |
| : 1. Isolation devices in high radiation areas may be verified by use of administrative means.
| |
| : 2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.
| |
| Verify the affected Once per 31 days penetration flow path following isolation is isolated.
| |
| D. MSIV leakage rate D.1 Restore leakage rate 8 hours not within limit. to within limit.
| |
| E. Required Action and E.1 Be in MODE 3. 12 hours associated Completion Time of Condition A, AND B, C, or D not met.
| |
| E.2 Be in MODE 4. 36 hours delete Quad Cities 1 and 2 3.6.1.3-4 Amendment No. 287/283
| |
| | |
| Low Set Relief Valves 3.6.1.6 3.6 CONTAINMENT SYSTEMS 3.6.1.6 Low Set Relief Valves LCO 3.6.1.6 The low set relief function of two relief valves shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One low set relief A.1 Restore low set 14 days valve inoperable. relief valve to OPERABLE status. Insert B here B. Required Action and --------------NOTE-----------
| |
| associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.
| |
| not met. -----------------------------
| |
| B.1 Be in MODE 3. 12 hours C. Two low set relief C.1 Be in MODE 3. 12 hours valves inoperable.
| |
| AND C.2 Be in MODE 4. 36 hours delete Quad Cities 1 and 2 3.6.1.6-1 Amendment No. 245/240
| |
| | |
| Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.7 3.6 CONTAINMENT SYSTEMS 3.6.1.7 Reactor Building-to-Suppression Chamber Vacuum Breakers LCO 3.6.1.7 Each reactor building-to-suppression chamber vacuum breaker shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS
| |
| -------------------------------------NOTE-------------------------------------
| |
| Separate Condition entry is allowed for each line.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more lines with A.1 Close the open vacuum 7 days one reactor building- breaker.
| |
| to-suppression chamber Insert B here vacuum breaker not closed.
| |
| B. One or more lines with B.1 Close one open vacuum 1 hour two reactor building- breaker.
| |
| to-suppression chamber vacuum breakers not closed.
| |
| C. One line with one or C.1 Restore the vacuum 7 days more reactor building- breaker(s) to to-suppression chamber OPERABLE status. Insert B here vacuum breakers inoperable for opening.
| |
| (continued)
| |
| Quad Cities 1 and 2 3.6.1.7-1 Amendment No. 199/195
| |
| | |
| Reactor Building-to-Suppression Chamber Vacuum Breakers 3.6.1.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and --------------NOTE-----------
| |
| Associated Completion LCO 3.0.4.a is not applicable Time of Condition C not when entering MODE 3.
| |
| met. -----------------------------
| |
| D.1 Be in MODE 3. 12 hours E. Two lines with one or E.1 Restore all vacuum 1 hour more reactor building- breakers in one line to-suppression chamber to OPERABLE status. Insert C here vacuum breakers inoperable for opening.
| |
| F. Required Action and F.1 Be in MODE 3. 12 hours Associated Completion Time of Conditions A, AND B or E not met.
| |
| F.2 Be in MODE 4. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1.7.1 ------------------NOTES------------------
| |
| : 1. Not required to be met for vacuum breakers that are open during Surveillances.
| |
| : 2. Not required to be met for vacuum breakers open when performing their intended function. delete Verify each vacuum breaker is closed. In accordance with the Surveillance Frequency Control Program (continued)
| |
| Quad Cities 1 and 2 3.6.1.7-2 Amendment No. 248/243
| |
| | |
| Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.8 3.6 CONTAINMENT SYSTEMS 3.6.1.8 Suppression Chamber-to-Drywell Vacuum Breakers LCO 3.6.1.8 Nine suppression chamber-to-drywell vacuum breakers shall be OPERABLE for opening.
| |
| AND Twelve suppression chamber-to-drywell vacuum breakers shall be closed.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required A.1 Restore one vacuum 72 hours suppression chamber- breaker to OPERABLE to-drywell vacuum status.
| |
| Insert B here breaker inoperable for opening.
| |
| B. Required Action and --------------NOTE-----------
| |
| associated Completion LCO 3.0.4.a is not applicable Time of Condition A when entering MODE 3.
| |
| not met. -----------------------------
| |
| B.1 Be in MODE 3. 12 hours delete C. One suppression C.1 Close the open vacuum 4 hours chamber-to-drywell breaker.
| |
| vacuum breaker not closed.
| |
| D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time of Condition C AND not met.
| |
| D.2 Be in MODE 4. 36 hours Quad Cities 1 and 2 3.6.1.8-1 Amendment No. 245/240
| |
| | |
| RHR Suppression Pool Cooling 3.6.2.3 3.6 CONTAINMENT SYSTEMS 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling LCO 3.6.2.3 Two RHR suppression pool cooling subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR suppression A.1 Restore RHR 7 days pool cooling subsystem suppression pool inoperable. cooling subsystem to Insert B here OPERABLE status.
| |
| B. Required Action and --------------NOTE-----------
| |
| associated Completion LCO 3.0.4.a is not applicable delete Time of Condition A when entering MODE 3.
| |
| not met. -----------------------------
| |
| B.1 Be in MODE 3. 12 hours C. Two RHR suppression C.1 Restore one RHR 8 hours pool cooling suppression pool subsystems inoperable. cooling subsystem to OPERABLE status.
| |
| D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time of Condition C AND not met.
| |
| D.2 Be in MODE 4. 36 hours Quad Cities 1 and 2 3.6.2.3-1 Amendment No. 245/240
| |
| | |
| RHR Drywell Spray 3.6.2.6 3.6 CONTAINMENT SYSTEMS delete 3.6.2.6 Residual Heat Removal (RHR) Drywell Spray LCO 3.6.2.6 Two RHR drywell spray subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR drywell spray A.1 Restore RHR drywell 7 days subsystem inoperable. spray subsystem to OPERABLE status. Insert B here B. Two RHR drywell spray B.1 Restore one RHR drywell 8 hours subsystems inoperable. spray subsystem to OPERABLE status.
| |
| C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours Quad Cities 1 and 2 3.6.2.6-1 Amendment No. 281/277
| |
| | |
| RHRSW System 3.7.1 3.7 PLANT SYSTEMS 3.7.1 Residual Heat Removal Service Water (RHRSW) System LCO 3.7.1 Two RHRSW subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHRSW pump A.1 Restore RHRSW pump to 30 days inoperable. OPERABLE status.
| |
| B. One RHRSW pump in each B.1 Restore one RHRSW 7 days subsystem inoperable. pump to OPERABLE status. Insert B here C. One RHRSW subsystem C.1 --------NOTE--------
| |
| inoperable for reasons Enter applicable other than Conditions and Condition A. Required Actions of LCO 3.4.7, "Residual Heat Removal (RHR)
| |
| Shutdown Cooling SystemHot Shutdown," for RHR shutdown cooling subsystem made inoperable by RHRSW System.
| |
| Restore RHRSW 7 days subsystem to OPERABLE status. Insert B here (continued)
| |
| Quad Cities 1 and 2 3.7.1-1 Amendment No. 199/195
| |
| | |
| SSMP System 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Safe Shutdown Makeup Pump (SSMP) System LCO 3.7.9 The SSMP System shall be OPERABLE.
| |
| APPLICABILITY: MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SSMP System A.1 Restore SSMP System 14 days inoperable. to OPERABLE status.
| |
| Insert B here B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. AND B.2 Reduce reactor steam 36 hours dome pressure to 150 psig.
| |
| SURVEILLANCE REQUIREMENTS delete SURVEILLANCE FREQUENCY SR 3.7.9.1 Verify each SSMP System manual, power In accordance operated, and automatic valve in the flow with the path, that is not locked, sealed, or Surveillance otherwise secured in position, is in the Frequency correct position. Control Program (continued)
| |
| Quad Cities 1 and 2 3.7.9-1 Amendment No. 248/2435
| |
| | |
| AC SourcesOperating 3.8.1 ACTIONS
| |
| -------------------------------------NOTE-------------------------------------
| |
| LCO 3.0.4.b is not applicable to the unit and common DGs, but is applicable to the opposite unit DG.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite A.1 Perform SR 3.8.1.1 1 hour circuit inoperable. for OPERABLE required offsite circuit. AND Once per 8 hours thereafter AND A.2 Declare required 24 hours from feature(s) with no discovery of no offsite power offsite power to available inoperable one division when the redundant concurrent with required feature(s) inoperability of are inoperable. redundant required feature(s)
| |
| AND delete A.3 Restore required 7 days offsite circuit to OPERABLE status. Insert B here (continued)
| |
| Quad Cities 1 and 2 3.8.1-2 Amendment No. 275/270
| |
| | |
| AC SourcesOperating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One required DG B.1 Perform SR 3.8.1.1 1 hour inoperable. for OPERABLE required offsite circuit(s). AND Once per 8 hours thereafter AND B.2 Declare required 4 hours from feature(s), supported discovery of by the inoperable DG, Condition B inoperable when the concurrent with redundant required inoperability of feature(s) are redundant inoperable. required feature(s)
| |
| AND B.3.1 Determine OPERABLE 24 hours DG(s) are not inoperable due to common cause failure.
| |
| OR B.3.2 Perform SR 3.8.1.2 24 hours for OPERABLE DG(s).
| |
| AND delete B.4 Restore required DG 7 days to OPERABLE status.
| |
| Insert B here (continued)
| |
| Quad Cities 1 and 2 3.8.1-3 Amendment No. 275/270
| |
| | |
| AC SourcesOperating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Two required offsite C.1 Declare required 12 hours from circuits inoperable. feature(s) inoperable discovery of when the redundant Condition C required feature(s) concurrent with are inoperable. inoperability of redundant required feature(s)
| |
| AND C.2 Restore one required 24 hours offsite circuit to OPERABLE status. Insert B here D. One required offsite ------------NOTE------------
| |
| circuit inoperable. Enter applicable Conditions and Required Actions of AND LCO 3.8.7, "Distribution SystemsOperating," when One required DG Condition D is entered with inoperable. no AC power source to any division.
| |
| D.1 Restore required 12 hours offsite circuit to OPERABLE status. Insert B here OR D.2 Restore required DG 12 hours to OPERABLE status.
| |
| Insert B here E. Two required DGs E.1 Restore one required 2 hours inoperable. DG to OPERABLE status.
| |
| (continued)
| |
| Quad Cities 1 and 2 3.8.1-4 Amendment No. 199/195
| |
| | |
| DC SourcesOperating 3.8.4 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC SourcesOperating LCO 3.8.4 The following DC electrical power subsystems shall be OPERABLE:
| |
| : a. Two 250 VDC electrical power subsystems;
| |
| : b. Division 1 and Division 2 125 VDC electrical power subsystems; and
| |
| : c. The opposite unit's 125 VDC electrical power subsystem capable of supporting equipment required to be OPERABLE by LCO 3.6.4.3, "Standby Gas Treatment (SGT) System,"
| |
| LCO 3.7.4, "Control Room Emergency Ventilation (CREV)
| |
| System" (Unit 2 only), LCO 3.7.5 "Control Room Emergency Ventilation Air Conditioning (AC) System" (Unit 2 only),
| |
| and LCO 3.8.1, "AC SourcesOperating."
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One 250 VDC electrical A.1 Restore the 250 VDC 72 hours power subsystem electrical power inoperable. subsystem to OPERABLE Insert B here status.
| |
| (continued)
| |
| Quad Cities 1 and 2 3.8.4-1 Amendment No. 199/195
| |
| | |
| DC SourcesOperating 3.8.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. ---------NOTE--------- B.1 Place associated 72 hours Only applicable if the OPERABLE alternate opposite unit is in 125 VDC electrical MODE 1, 2, or 3. power subsystem in
| |
| ---------------------- service.
| |
| Division 1 or 2 AND 125 VDC battery inoperable as a result B.2 Restore Division 1 or Prior to of maintenance or 2 125 VDC battery to exceeding 7 testing. OPERABLE status. cumulative days per operating cycle of battery inoperability, on a per battery basis, as a result of maintenance or testing Insert B here C. ---------NOTE--------- C.1 Place associated 72 hours Only applicable if the OPERABLE alternate opposite unit is in 125 VDC electrical MODE 1, 2, or 3. power subsystem in
| |
| ---------------------- service.
| |
| Division 1 or 2 AND 125 VDC battery inoperable, due to the C.2 Restore Division 1 or 7 days need to replace the 2 125 VDC battery to battery, as determined OPERABLE status.
| |
| Insert B here by maintenance or testing.
| |
| (continued)
| |
| Quad Cities 1 and 2 3.8.4-2 Amendment No. 199/195
| |
| | |
| DC SourcesOperating 3.8.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Division 1 or 2 D.1 Restore Division 1 or 72 hours 125 VDC electrical 2 125 VDC electrical power subsystem power subsystem to Insert B here inoperable for reasons OPERABLE status.
| |
| other than Conditions B or C. OR D.2 --------NOTE---------
| |
| Only applicable if the opposite unit is not in MODE 1, 2, or 3.
| |
| Place associated 72 hours OPERABLE alternate 125 VDC electrical power subsystem in service.
| |
| E. Opposite unit 125 VDC E.1 Restore the opposite 7 days electrical power unit 125 VDC subsystem inoperable. electrical power Insert B here subsystem to OPERABLE status.
| |
| F. Required Action and --------------NOTE-----------
| |
| associated Completion LCO 3.0.4.a is not applicable delete Time not met. when entering MODE 3.
| |
| F.1 Be in MODE 3. 12 hours Quad Cities 1 and 2 3.8.4-3 Amendment No. 245/240
| |
| | |
| Distribution SystemsOperating 3.8.7 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Distribution SystemsOperating LCO 3.8.7 The following electrical power distribution subsystems shall be OPERABLE:
| |
| : a. Division 1 and Division 2 AC and DC electrical power distribution subsystems; and
| |
| : b. The portions of the opposite unit's AC and DC electrical power distribution subsystems necessary to support equipment required to be OPERABLE by LCO 3.6.4.3, "Standby Gas Treatment (SGT) System," LCO 3.7.4, "Control Room Emergency Ventilation (CREV) System" (Unit 2 only), LCO 3.7.5, "Control Room Emergency Ventilation Air Conditioning (AC) System" (Unit 2 only),
| |
| and LCO 3.8.1, "AC SourcesOperating."
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME delete A. One or more AC A.1 Restore AC electrical 8 hours electrical power power distribution distribution subsystems to Insert C here subsystems inoperable. OPERABLE status.
| |
| (continued)
| |
| Quad Cities 1 and 2 3.8.7-1 Amendment No. 275/270
| |
| | |
| Distribution SystemsOperating 3.8.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME delete B. One or more DC B.1 Restore DC electrical 2 hours electrical power power distribution distribution subsystems to Insert B here subsystems inoperable. OPERABLE status.
| |
| C. One or more required -------------NOTE------------
| |
| opposite unit AC or DC Enter applicable Condition electrical power and Required Actions of distribution LCO 3.8.1 when Condition C subsystems inoperable. results in the inoperability of a required offsite circuit.
| |
| C.1 Restore required 7 days opposite unit AC and DC electrical power Insert B here distribution subsystems to OPERABLE status.
| |
| D. Required Action and --------------NOTE-----------
| |
| associated Completion LCO 3.0.4.a is not applicable Time of Condition A, when entering MODE 3.
| |
| B, or C not met. -----------------------------
| |
| D.1 Be in MODE 3. 12 hours E. Two or more electrical E.1 Enter LCO 3.0.3. Immediately power distribution subsystems inoperable that, in combination, result in a loss of function.
| |
| Quad Cities 1 and 2 3.8.7-2 Amendment No. 275/270
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
| |
| delete
| |
| : a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.
| |
| : b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, Risk-Informed Method for Control of Surveillance Frequencies, Revision 1.
| |
| : c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
| |
| Insert E Quad Cities 1 and 2 5.5-14 Amendment No. 248/243
| |
| | |
| Insert A Example 1.3-8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Restore subsystem 7 days subsystem to OPERABLE inoperable. status. OR In accordance with the Risk Informed Completion Time Program B. Required B.1 Be in MODE 3. 12 hours Action and associated AND Completion Time not B.2 Be in MODE 4. 36 hours met.
| |
| When a subsystem is declared inoperable, Condition A is entered. The 7 day Completion Time may be applied as discussed in Example 1.3-2. However, the licensee may elect to apply the Risk Informed Completion Time Program which permits calculation of a Risk Informed Completion Time (RICT) that may be used to complete the Required Action beyond the 7 day Completion Time. The RICT cannot exceed 30 days. After the 7 day Completion Time has expired, the subsystem must be restored to OPERABLE status within the RICT or Condition C must also be entered.
| |
| The Risk Informed Completion Time Program requires recalculation of the RICT to reflect changing plant conditions. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours after the plant configuration change, whichever is less.
| |
| If the 7 day Completion Time clock of Condition A has expired and subsequent changes in plant condition result in exiting the applicability of the Risk Informed Completion Time Program without restoring the inoperable subsystem to
| |
| | |
| OPERABLE status, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start.
| |
| If the RICT expires or is recalculated to be less than the elapsed time since the Condition was entered and the inoperable subsystem has not been restored to OPERABLE status, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable subsystems are restored to OPERABLE status after Condition B is entered, Condition A is exited, and therefore, the Required Actions of Condition B may be terminated.
| |
| | |
| Insert B OR In accordance with the Risk Informed Completion Time Program Insert C OR
| |
| --------NOTE--------
| |
| Only applicable when a loss of function has not occurred.
| |
| In accordance with the Risk Informed Completion Time Program Insert D OR
| |
| --------NOTE--------
| |
| Not applicable if leakage exceeds limits or if loss of function has occurred.
| |
| In accordance with the Risk Informed Completion Time Program Insert E 5.5.15 Risk Informed Completion Time Program This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines." The program shall include the following:
| |
| : a. The RICT may not exceed 30 days;
| |
| : b. A RICT may only be utilized in MODE 1 and 2;
| |
| : c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.
| |
| : 1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.
| |
| : 2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours after the plant configuration change, whichever is less.
| |
| : 3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.
| |
| : d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:
| |
| : 1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or
| |
| : 2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.
| |
| : e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods used to support this license amendment, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.
| |
| | |
| ATTACHMENT 3 Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Proposed Technical Specification Bases Changes (Mark-Ups)
| |
| (For Information Only)
| |
| TS Bases Pages B 3.1.7-3 B 3.6.1.3-5 and -6 B 3.3.1.1-23 and -25 B 3.6.1.3-8 B 3.3.2.2-4 B 3.6.1.6-2 B 3.3.4.1-6 B 3.6.1.7-4 thru -6 B 3.3.5.1-31 B 3.6.1.8-4 B 3.3.5.1-33 thru -35 B 3.6.2.3-2 B 3.3.5.1-37 and -39 B 3.6.2.6-2 B 3.3.5.3-8 and -10 B 3.7.1-4 B 3.3.6.1-21 B 3.7.9-2 B 3.3.6.3-4 B 3.8.1-9 and -12 B 3.3.8.1-5 B 3.8.1-15 and -16 B 3.4.3-4 B 3.8.4-5 thru -7 B 3.5.1-7 B 3.8.4-9 B 3.5.1-9 and -10 B 3.8.7-5 and -7 B 3.5.3-3 B 3.8.7-9 B 3.6.1.2-6
| |
| | |
| SLC System B 3.1.7 BASES (continued)
| |
| APPLICABILITY In MODES 1 and 2, shutdown capability is required. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical. In MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Demonstration of adequate SDM (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") ensures that the reactor will not become critical. Therefore, the SLC System is not required to be OPERABLE when only a single control rod can be withdrawn.
| |
| In MODES 1, 2, and 3, the SLC System must be OPERABLE to ensure that offsite doses remain within 10 CFR 50.67 (Ref.
| |
| : 4) limits following a LOCA involving significant fission product releases. The SLC System is designed to maintain suppression pool pH at or above 7 following a LOCA to ensure that iodine will be retained in the suppression pool water (Ref. 3).
| |
| or in accordance with the Risk Informed ACTIONS A.1 Completion Time Program If one SLC subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE subsystem is adequate to shutdown the unit and meet the requirement of Reference 1.
| |
| However, the overall capability is reduced because a single failure in the remaining OPERABLE subsystem could result in reduced SLC System shutdown capability. The 7 day Completion Time is based on the availability of an OPERABLE subsystem capable of shutting down the reactor and the low probability of a Design Basis Accident (DBA) or severe deleted transient occurring concurrent with the failure of the Control Rod Drive (CRD) System to shut down the reactor.
| |
| B.1 If both SLC subsystems are inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours. The allowed Completion Time of 8 hours is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.1.7-3 Revision 33
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES ACTIONS entry into the Condition. However, the Required Actions for (continued) inoperable RPS instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel.
| |
| Note 2 has been provided to modify the ACTIONS for the RPS instrumentation functions of APRM Flow Biased Neutron Flux High (Function 2.b) and APRM Fixed Neutron FluxHigh (Function 2.c) when they are inoperable due to failure of SR 3.3.1.1.2 and gain adjustments are necessary. Note 2 allows entry into associated Conditions and Required Actions deleted to be delayed for up to 2 hours if the APRM is indicating a lower power value than the calculated power (i.e., the gain adjustment factor (GAF) is high (non-conservative)). The GAF for any channel is defined as the power value determined by the heat balance divided by the APRM reading for that channel. Upon completion of the gain adjustment, or expiration of the allowed time, the channel must be returned to OPERABLE status or the applicable Condition entered and the Required Actions taken. This Note is based on the time required to perform gain adjustments on multiple channels.
| |
| A.1 and A.2 Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours has been shown to be acceptable (Ref. 13) to permit restoration of any inoperable required channel to OPERABLE status. However, this out of service time is only acceptable provided the associated Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to Required Actions B.1, B.2, and C.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel or the associated trip system must be placed in the tripped condition per Required Actions A.1 and A.2. Placing the inoperable channel in trip (or the associated trip system in (continued)
| |
| Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. A Note has been provided to indicate that a RICT is not applicable when a loss of safety function has occurred.
| |
| Quad Cities 1 and 2 B 3.3.1.1-23 Revision 62
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES ACTIONS B.1 and B.2 (continued) inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e., what MODE the plant is in).
| |
| If this action would result in a scram, it is permissible to place the other trip system or its inoperable channels in trip.
| |
| The 6 hour Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.
| |
| Alternatively, a Alternately, if it is not desired to place the inoperable Completion Time channels (or one trip system) in trip (e.g., as in the case can be determined where placing the inoperable channel or associated trip in accordance with system in trip would result in a scram), Condition D must be the Risk Informed entered and its Required Action taken. The 6 hour allowance Completion Time is not allowed for Reactor Mode SwitchShutdown and Manual Program. A Note Scram Function channels since with two channels inoperable RPS trip capability is not maintained. In this case, has been provided Condition C must be entered and its Required Action taken.
| |
| to indicate that a RICT is not applicable when a C.1 loss of safety function has Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped occurred. channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability.
| |
| A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal. For the typical Function with one-out-of-two taken twice logic and the IRM and APRM Functions, this would require both trip systems to have one channel OPERABLE or in trip (or the associated trip system in trip). For Function 5 (Main Steam Isolation Valve-Closure), this would require both trip systems to (continued)
| |
| Quad Cities 1 and 2 B 3.3.1.1-25 Revision 0
| |
| | |
| Feedwater System and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES ACTIONS A.1 (continued) deleted System and main turbine high water level trip capability maintained (refer to Required Action B.1 Bases), the Feedwater System and Main Turbine High Water Level Trip Instrumentation is capable of performing the intended function. However, the reliability and redundancy of the Feedwater System and Main Turbine High Water Level Trip Instrumentation is reduced, such that a single failure in one of the remaining channels, concurrent with feedwater controller failure, maximum demand event, may result in the inability of the instrumentation to perform the intended function. Therefore, continued operation is only allowed for a limited time. If the inoperable channel(s) cannot be restored to OPERABLE status within the Completion Time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel(s) in trip would conservatively compensate for the inoperability(s), restore capability to accommodate a single failure, and allow Alternatively, a operation to continue with no further restrictions. As noted, placing the channel in trip with no further Completion Time can restrictions is not allowed if the inoperable channel is the be determined in result of an inoperable feedwater pump breaker, since this accordance with the may not adequately compensate for the inoperable breaker.
| |
| Risk Informed Alternatively, if it is not desired to place the channel(s)
| |
| Completion Time in trip (e.g., as in the case where placing the inoperable Program. A Note has channel(s) in trip would result in feedwater pump trip and main turbine trip), or if the inoperable channel(s) is the been provided to result of an inoperable breaker, Condition C must be entered indicate that a RICT is and its Required Actions taken.
| |
| not applicable when a loss of safety function The Completion Time of 7 days is based on the low probability of the event occurring coincident with a single has occurred.
| |
| failure in a remaining OPERABLE channel.
| |
| B.1 With the Feedwater System and main turbine high water level trip capability not maintained, the Feedwater System and Main Turbine High Water Level Trip Instrumentation cannot perform its design function. Therefore, continued operation is only permitted for a 2 hour period, during which Feedwater System and main turbine high water level trip capability must be restored. The trip capability is considered maintained when sufficient channels are OPERABLE or in trip such that the Feedwater System and main turbine high water level trip logic will generate a trip signal on a (continued)
| |
| Quad Cities 1 and 2 B 3.3.2.2-4 Revision 32
| |
| | |
| ATWS-RPT Instrumentation B 3.3.4.1 BASES ACTIONS A.1 and A.2 (continued) reliability and redundancy of the ATWS-RPT instrumentation is reduced, such that a single failure in the remaining trip system could result in the inability of the ATWS-RPT System to perform the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to OPERABLE status. Because of the diversity of sensors available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of ATWS-RPT, 14 days is provided to restore the inoperable channel (Required Action A.1). Alternately, the inoperable channel may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a deleted single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an Alternatively, a inoperable ASD feed breaker or ASD emergency stop circuit, Completion Time since this may not adequately compensate for the inoperable ASD feed breaker or ASD emergency stop circuit (e.g., the can be determined device may be inoperable such that it will not provide in accordance with tripping capability). If it is not desired to place the the Risk Informed channel in trip (e.g., as in the case where placing the Completion Time inoperable channel in trip would result in an RPT), or if Program. A Note the inoperable channel is the result of an inoperable has been provided breaker, Condition D must be entered and its Required Actions taken.
| |
| to indicate that a RICT is not applicable when a B.1 loss of safety Required Action B.1 is intended to ensure that appropriate function has actions are taken if multiple, inoperable, untripped occurred. channels within the same Function result in the Function not maintaining ATWS-RPT trip capability. A Function is considered to be maintaining ATWS-RPT trip capability when sufficient channels are OPERABLE or in trip such that the ATWS-RPT System will generate a trip signal from the given Function on a valid signal, and both recirculation pumps can be tripped. This requires two channels of the Function in the same trip system to each be OPERABLE or in trip, and the recirculation pump ASD feed breakers and ASD emergency stop circuitry to be OPERABLE or in trip.
| |
| (continued) deleted Quad Cities 1 and 2 B 3.3.4.1-6 Revision 41
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS B.1, B.2, and B.3 (continued) upon discovery that a redundant feature in the same system (e.g., both CS subsystems) cannot be automatically initiated due to inoperable, untripped channels within the same Function as described in the paragraph above. For Required Action B.2, the Completion Time only begins upon discovery that the HPCI System cannot be automatically initiated due to two inoperable, untripped channels for the associated variable in the same trip system. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
| |
| Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to Alternatively, a be acceptable (Ref. 4) to permit restoration of any Completion Time inoperable channel to OPERABLE status. If the inoperable can be determined channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in in accordance with the tripped condition per Required Action B.3. Placing the the Risk Informed inoperable channel in trip would conservatively compensate Completion Time for the inoperability, restore capability to accommodate a Program. A Note single failure, and allow operation to continue.
| |
| has been provided Alternately, if it is not desired to place the channel in to indicate that a trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition H RICT is not must be entered and its Required Action taken.
| |
| applicable when a loss of safety function has C.1 and C.2 occurred.
| |
| Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the same variable result in redundant automatic initiation capability being lost for the feature(s). Required Action C.1 features would be those that are initiated by Functions 1.c, 1.e, 2.c, 2.e, 2.g, 2.h, 2.i, and 2.k (i.e.,
| |
| low pressure ECCS). Redundant automatic initiation capability is lost if either (a) two Function 1.c channels are inoperable in both trip systems, (b) two Function 2.c channels are inoperable in both trip systems, (c) two Function 1.e channels are inoperable, (d) two Function 2.e channels are inoperable, (e) two or more Function 2.g (continued)
| |
| Quad Cities 1 and 2 B 3.3.5.1-31 Revision 0
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS C.1 and C.2 (continued) this Condition if a channel in this Function is inoperable),
| |
| since the loss of the Function was considered during the development of Reference 4 and considered acceptable for the 24 hours allowed by Required Action C.2.
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
| |
| For Required Action C.1, the Completion Time only begins upon discovery that the same feature in both subsystems (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same variable as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.
| |
| Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would either cause the initiation or it would not necessarily Alternatively, a result in a safe state for the channel in all events.
| |
| Completion Time can be determined D.1, D.2.1, and D.2.2 in accordance with the Risk Informed Required Action D.1 is intended to ensure that appropriate Completion Time actions are taken if multiple, inoperable, untripped Program. A Note channels within the same Function result in a complete loss has been provided of automatic component initiation capability for the HPCI System. If both CCSTs are available, HPCI automatic to indicate that a initiation capability is lost if two required Function 3.d RICT is not channels are inoperable and untripped. If one CCST is not applicable when a available, automatic initiation capability is lost if two loss of safety channels associated with the aligned CCST are inoperable and function has untripped. HPCI automatic initiation capability is lost if occurred.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.3.5.1-33 Revision 0
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS D.1, D.2.1, and D.2.2 (continued) two Function 3.e channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour allowance of Required Actions D.2.1 and D.2.2 is not appropriate and the HPCI System must be declared inoperable within 1 hour after discovery of loss of HPCI initiation capability. As noted, Required Action D.1 is only applicable if the HPCI pump suction is not aligned to the suppression pool, since, if aligned, the Function is already performed.
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
| |
| For Required Action D.1, the Completion Time only begins upon discovery that the HPCI System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
| |
| Alternatively, a Because of the diversity of sensors available to provide Completion Time initiation signals and the redundancy of the ECCS design, an can be determined allowable out of service time of 24 hours has been shown to in accordance with be acceptable (Ref. 4) to permit restoration of any within 24 the Risk Informed inoperable channel to OPERABLE status. If the inoperable hours Completion Time channel cannot be restored to OPERABLE status within the Program. A Note allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1 or the has been provided suction source must be aligned to the suppression pool per to indicate that a Required Action D.2.2. Placing the inoperable channel in RICT is not trip performs the intended function of the channel (shifting applicable when a the suction source to the suppression pool). Performance of loss of safety either of these two Required Actions will allow operation to continue. If Required Action D.2.1 or D.2.2 is performed, function has measures should be taken to ensure that the HPCI System occurred. piping remains filled with water. Alternately, if it is not desired to perform Required Actions D.2.1 and D.2.2 (e.g.,
| |
| as in the case where shifting the suction source could drain down the HPCI suction piping), Condition H must be entered and its Required Action taken.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.3.5.1-34 Revision 0
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS E.1 and E.2 (continued)
| |
| Required Action E.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the Core Spray and Low Pressure Coolant Injection Pump Discharge FlowLow (Bypass) Functions result in redundant automatic initiation capability being lost for the feature(s). For Required Action E.1, the features would be those that are initiated by Functions 1.d and 2.f (i.e., low pressure ECCS). Redundant automatic initiation capability is lost if (a) two Function 1.d channels are inoperable or (b) two Function 2.f channels are inoperable. Since each inoperable channel would have Required Action E.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected low pressure ECCS pump to be declared inoperable. However, since channels for more than one low pressure ECCS pump are inoperable, and the Completion Times started concurrently for the channels of the low pressure ECCS pumps, this results in the affected low pressure ECCS pumps being concurrently declared inoperable.
| |
| In this situation (loss of redundant automatic initiation capability), the 7 day allowance of Required Action E.2 is deleted not appropriate and the subsystem associated with each inoperable channel must be declared inoperable within 1 hour. A Note is also provided (the Note to Required Action E.1) to delineate that Required Action E.1 is only applicable to low pressure ECCS Functions. Required Action E.1 is not applicable to HPCI Function 3.f since the loss of one channel results in a loss of the Function (one-out-of-one logic). This loss was considered during the development of Reference 4 and considered acceptable for the 7 days allowed by Required Action E.2.
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal (continued)
| |
| Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. A Note has been provided to indicate that a RICT is not applicable when a loss of safety function has occurred.
| |
| Quad Cities 1 and 2 B 3.3.5.1-35 Revision 61
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS F.1 and F.2 (continued)
| |
| Required Action F.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within similar ADS trip system A and B Functions result in redundant automatic initiation capability being lost for the ADS. Redundant automatic initiation capability is lost if either (a) one or more Function 4.a channels and one or more Function 5.a channels are inoperable and untripped or (b) one or more Function 4.b channels and one or more Function 5.b channels are inoperable and untripped.
| |
| In this situation (loss of automatic initiation capability),
| |
| the 96 hour or 8 day allowance, as applicable, of Required Action F.2 is not appropriate and all ADS valves must be declared inoperable within 1 hour after discovery of loss of ADS initiation capability.
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
| |
| For Required Action F.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable, untripped channels within similar ADS trip system Functions as described in the paragraph above. The 1 hour Completion Time from discovery Alternatively, a of loss of initiation capability is acceptable because it Completion Time minimizes risk while allowing time for restoration or tripping of channels.
| |
| can be determined in accordance with Because of the diversity of sensors available to provide the Risk Informed initiation signals and the redundancy of the ECCS design, an Completion Time allowable out of service time of 8 days has been shown to be Program. A Note acceptable (Ref. 4) to permit restoration of any inoperable has been provided channel to OPERABLE status if both HPCI and RCIC are OPERABLE. If either HPCI or RCIC is inoperable, the time is to indicate that a shortened to 96 hours. If the status of HPCI or RCIC RICT is not changes such that the Completion Time changes from 8 days to applicable when a 96 hours, the 96 hours begins upon discovery of HPCI or RCIC loss of safety inoperability. However, the total time for an inoperable, function has untripped channel cannot exceed 8 days. If the status of HPCI or RCIC changes such that the Completion Time changes occurred.
| |
| (continued) or in accordance with the Risk Informed Completion Time Program Quad Cities 1 and 2 B 3.3.5.1-37 Revision 0
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS G.1 and G.2 (continued) initiated due to inoperable channels within similar ADS trip system Functions as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
| |
| Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an Alternatively, a allowable out of service time of 8 days has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable Completion Time channel to OPERABLE status if both HPCI and RCIC are can be determined OPERABLE (Required Action G.2). If either HPCI or RCIC is in accordance with inoperable, the time shortens to 96 hours. If the status of the Risk Informed HPCI or RCIC changes such that the Completion Time changes Completion Time from 8 days to 96 hours, the 96 hours begins upon discovery Program. A Note of HPCI or RCIC inoperability. However, the total time for an inoperable channel cannot exceed 8 days. If the status has been provided of HPCI or RCIC changes such that the Completion Time to indicate that a changes from 96 hours to 8 days, the "time zero" for RICT is not beginning the 8 day "clock" begins upon discovery of the applicable when a inoperable channel. If the inoperable channel cannot be loss of safety restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required function has Action taken. The Required Actions do not allow placing the occurred. channel in trip since this action would not necessarily result in a safe state for the channel in all events.
| |
| or in accordance with the Risk Informed H.1 Completion Time Program With any Required Action and associated Completion Time not met, the associated feature(s) may be incapable of performing the intended function, and the supported feature(s) associated with inoperable untripped channels must be declared inoperable immediately.
| |
| SURVEILLANCE As noted in the beginning of the SRs, the SRs for each ECCS REQUIREMENTS instrumentation Function are found in the SRs column of Table 3.3.5.1-1.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.3.5.1-39 Revision 0
| |
| | |
| RCIC System Instrumentation B 3.3.5.3 deleted BASES ACTIONS B.1 and B.2 (continued)
| |
| Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic initiation capability for the RCIC System. In this case, automatic initiation capability is lost if two Function 1 channels in the same trip system are inoperable and untripped. In this situation (loss of automatic initiation capability), the 24 hour allowance of Required Action B.2 is not appropriate, and the RCIC System must be declared inoperable within 1 hour after discovery of loss of RCIC initiation capability.
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
| |
| For Required Action B.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically initiated due to two inoperable, untripped Reactor Vessel Alternatively, a Water LevelLow Low channels in the same trip system. The Completion Time can be 1 hour Completion Time from discovery of loss of initiation determined in capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
| |
| accordance with the Risk Informed Completion Because of the redundancy of sensors available to provide Time Program. A Note initiation signals and the fact that the RCIC System is not has been provided to credited in any accident or transient analysis, an allowable indicate that a RICT is out of service time of 24 hours has been shown to be not applicable when a acceptable (Ref. 1) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel loss of safety function cannot be restored to OPERABLE status within the allowable has occurred. out of service time, the channel must be placed in the tripped condition per Required Action B.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.
| |
| Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition E must be entered and its Required Action taken.
| |
| (continued) deleted Quad Cities 1 and 2 B 3.3.5.3-8 Revision 61
| |
| | |
| deleted RCIC System Instrumentation B 3.3.5.3 BASES ACTIONS D.1, D.2.1, and D.2.2 (continued)
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
| |
| For Required Action D.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
| |
| Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 1) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1, which performs within 24 the intended function of the channel (shifting the suction hours Alternatively, a source to the suppression pool). Alternatively, Required Completion Time Action D.2.2 allows the manual alignment of the RCIC suction can be determined to the suppression pool, which also performs the intended in accordance with function. If Required Action D.2.1 or D.2.2 is performed, the Risk Informed measures should be taken to ensure that the RCIC System piping remains filled with water. If it is not desired to Completion Time perform Required Actions D.2.1 and D.2.2 (e.g., as in the Program. A Note case where shifting the suction source could drain down the has been provided RCIC suction piping), Condition E must be entered and its to indicate that a Required Action taken.
| |
| RICT is not applicable when a E.1 loss of safety function has With any Required Action and associated Completion Time not occurred. met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared inoperable immediately.
| |
| (continued) deleted Quad Cities 1 and 2 B 3.3.5.3-10 Revision 61
| |
| | |
| Primary Containment Isolation Instrumentation B 3.3.6.1 BASES ACTIONS A.1 (continued)
| |
| Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours or 24 hours, depending on the Function (12 hours for those Functions that have channel components common to RPS instrumentation and 24 hours for those Functions that do not have channel components common to RPS instrumentation), has been shown to be acceptable (Refs. 9 and 10) to permit restoration of any inoperable channel to OPERABLE status.
| |
| This out of service time is only acceptable provided the Alternatively, a Completion associated Function is still maintaining isolation Time can be determined in capability (refer to Required Action B.1 Bases). If the accordance with the Risk inoperable channel cannot be restored to OPERABLE status Informed Completion Time within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1.
| |
| Program. A Note has been Placing the inoperable channel in trip would conservatively provided to indicate that a compensate for the inoperability, restore capability to RICT is not applicable when accommodate a single failure, and allow operation to a loss of safety function has continue with no further restrictions. Alternately, if it occurred. is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an isolation), Condition C must be entered and its Required Action taken.
| |
| B.1 Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in redundant automatic isolation capability being lost for the associated penetration flow path(s). The MSL and Primary Containment Isolation Functions and portions of other system Isolation Functions are considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip, such that both trip systems will generate a trip signal from the given Function on a valid signal. For Functions 1.a, 1.b, 1.c, 2.a, 2.b, 2.c, 5.b, 6.a, and 6.b, this would require both trip systems to have one channel OPERABLE or in trip. For Function 1.d, this would require both trip systems to have one channel, associated with each MSL, OPERABLE or in trip. Function 1.e, consists of channels (continued)
| |
| Quad Cities 1 and 2 B 3.3.6.1-21 Revision 23
| |
| | |
| Relief Valve Instrumentation B 3.3.6.3 BASES APPLICABILITY nuclear system and the relief valves may be needed to (continued) provide pressure relief. If the relief valves are needed, then the relief valve function is required to ensure that the primary containment design basis is maintained. In MODES 4 and 5, the reactor pressure is low enough that the overpressure limit cannot be approached by assumed operational transients or accidents. Thus, relief valve instrumentation and associated pressure relief is not required.
| |
| ACTIONS A.1 The failure of any relief valve instrument channel to provide the pressure setpoint or low set time delay for an individual relief valve does not affect the ability of the other relief valves to perform their relief or low set function. A relief valve is OPERABLE if the associated logic, has one Function 1.a or 2.a channel, as applicable, and, for low set relief valves, two Function 1.b channels OPERABLE. Therefore, 14 days is provided to restore the inoperable channel(s) to OPERABLE status (Required Action A.1). If the inoperable channel(s) cannot be restored to OPERABLE status within the allowable out of service time, Condition B must be entered and its Required Alternatively, a Completion Action taken. The 14 day Completion Time is considered appropriate because of the redundancy in the design (five Time can be determined in relief valves are provided and any four relief valves can accordance with the Risk perform the relief function, two low set relief valves are Informed Completion Time provided and one low set relief valve can perform the low Program. A Note has been set function) and the very low probability of multiple provided to indicate that a relief instrumentation channel failures, which render the RICT is not applicable remaining relief valves inoperable, occurring together with an event requiring the relief or low set function during the when a loss of safety 14 day Completion Time. The 14 day Completion Time to function has occurred. restore inoperable channels to OPERABLE status is based on the relief capability of the remaining relief valves, the low probability of an event requiring relief valve actuation and a reasonable time to complete the Required Action.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.3.6.3-4 Revision 0
| |
| | |
| LOP Instrumentation B 3.3.8.1 BASES APPLICABLE 2. 4160 V ESS Bus Undervoltage (Degraded Voltage)
| |
| SAFETY ANALYSES, (continued)
| |
| LCO, and APPLICABILITY allow restoration to normal voltages, but short enough to ensure that sufficient power is available to the required equipment.
| |
| Two channels of 4160 V ESS Bus Undervoltage/Time Delay (Degraded Voltage) Function and one channel of Degraded Voltage-Time Delay Function per associated bus are required to be OPERABLE when the associated DG is required to be deleted OPERABLE to ensure that no single instrument failure can preclude the degraded voltage and time delay function.
| |
| Refer to LCO 3.8.1 for Applicability Bases for the DGs.
| |
| ACTIONS A Note has been provided to modify the ACTIONS related to LOP instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Alternatively, a Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial Completion Time can be entry into the Condition. However, the Required Actions for determined in accordance inoperable LOP instrumentation channels provide appropriate with the Risk Informed compensatory measures for separate inoperable channels. As Completion Time such, a Note has been provided that allows separate Program. A Note has Condition entry for each inoperable LOP instrumentation been provided to indicate channel.
| |
| that a RICT is not applicable when a loss of A.1 safety function has occurred. With one or more channels of a Function inoperable, the Function is not capable of performing the intended function.
| |
| Therefore, only 1 hour is allowed to restore the inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel in trip would conservatively compensate (continued)
| |
| Quad Cities 1 and 2 B 3.3.8.1-5 Revision 740
| |
| | |
| Safety and Relief Valves B 3.4.3 BASES APPLICABILITY In MODES 1, 2, and 3, all safety and relief valves must be OPERABLE, since considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The safety and relief valves may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.
| |
| In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure and MCPR limits are unlikely to be approached by assumed operational transients or accidents.
| |
| In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The safety and relief functions are not needed during these conditions.
| |
| ACTIONS A.1 With the relief function of one relief valve (or S/RV) inoperable, the remaining OPERABLE relief valves are capable of providing the necessary protection. However, the overall reliability of the pressure relief system is reduced because additional failures in the remaining OPERABLE relief valves could result in failure to adequately relieve pressure during a limiting event. For this reason, continued operation is permitted for a limited time only.
| |
| The 14 day Completion Time to restore the inoperable required relief valve to OPERABLE status is based on the relief capability of the remaining relief valves, the low probability of an event requiring relief valve actuation, deleted and a reasonable time to complete the Required Action.
| |
| B.1 If the relief function of the inoperable relief valve or S/RV cannot be restored to OPERABLE status within the associated Completion Time of Required Action A.1, the plant must be brought to a MODE in which the overall plant risk is minimized. To achieve this status, the plant must be brought to MODE 3 within 12 hours. Remaining in the Applicability of Alternatively, a Completion Time (continued) can be determined in accordance with the Risk Informed Completion Time Program.
| |
| Quad Cities 1 and 2 B 3.4.3-4 Revision 40
| |
| | |
| ECCSOperating B 3.5.1 BASES ACTIONS A.1 (continued) reliability is reduced, because a single failure in one of the remaining OPERABLE LPCI subsystems, concurrent with a LOCA, may result in the LPCI subsystems not being able to perform their intended safety function. The 30 day Completion Time is based on a reliability study cited in Reference 10 that evaluated the impact on ECCS availability, assuming various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowable repair times (i.e., Completion Times).
| |
| or in accordance with the Risk B.1 Informed Completion Time Program If a LPCI subsystem is inoperable for reasons other than Condition A, or a CS subsystem is inoperable, the inoperable low pressure ECCS injection/spray subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced, because a single failure in one of the remaining OPERABLE subsystems, concurrent with a LOCA, may result in the ECCS not being able to perform its intended safety function. The 7 day Completion Time is based on a reliability study (Ref. 10) that evaluated the impact on ECCS availability, assuming various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowed outage times (i.e., Completion Times).
| |
| or in accordance with the Risk Informed C.1 Completion Time Program If one LPCI pump in each subsystem is inoperable, one LPCI pump must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE ECCS subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced because a single failure in one of the remaining OPERABLE ECCS subsystems, concurrent with a LOCA, may result in the ECCS not being able to perform its intended safety function. The 7 day Completion Time is based on a reliability study (Ref. 10) that (continued)
| |
| Quad Cities 1 and 2 B 3.5.1-7 Revision 22
| |
| | |
| ECCSOperating B 3.5.1 or in accordance with the Risk Informed BASES Completion Time Program deleted ACTIONS E.1 (continued)
| |
| If two LPCI subsystems are inoperable for reasons other than Condition C, one inoperable subsystem must be restored to OPERABLE status within 72 hours. In this Condition, the remaining OPERABLE CS subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced, because a single failure in one of the remaining CS subsystems, concurrent with a LOCA, may result in ECCS not being able to perform its intended safety function. The 72 hour Completion Time is based on a reliability study cited in Reference 10 that evaluated the impact on ECCS availability, assuming various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowable repair times (i.e., Completion Times).
| |
| deleted F.1 and F.2 If the Required Action and associated Completion Time of Condition E is not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| G.1 and G.2 If the HPCI System is inoperable and the RCIC System is verified to be OPERABLE, the HPCI System must be restored to OPERABLE status within 14 days. In this Condition, adequate core cooling is ensured by the OPERABILITY of the redundant or in accordance and diverse low pressure ECCS injection/spray subsystems in with the Risk conjunction with ADS. Also, the RCIC System will Informed automatically provide makeup water at most reactor operating Completion Time pressures. Verification of RCIC OPERABILITY is therefore Program required immediately when HPCI is inoperable. This may be performed as an administrative check by examining logs or other information to determine if RCIC is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate the OPERABILITY of (continued)
| |
| Quad Cities 1 and 2 B 3.5.1-9 Revision 40
| |
| | |
| ECCSOperating B 3.5.1 BASES ACTIONS G.1 and G.2 (continued) deleted the RCIC System. If the OPERABILITY of the RCIC System cannot be verified, however, Condition I must be immediately entered. In the event of component failures concurrent with a design basis LOCA, there is a potential, depending on the specific failures, that the minimum required ECCS equipment will not be available. A 14 day Completion Time is based on a reliability study cited in Reference 10 and has been found to be acceptable through operating experience.
| |
| H.1 The LCO requires five ADS valves to be OPERABLE in order to provide the ADS function. With one ADS valve out of service, the overall reliability of the ADS is reduced, because a single failure in the OPERABLE ADS valves could result in a reduction in depressurization capability.
| |
| Therefore, operation is only allowed for a limited time.
| |
| The 14 day Completion Time is based on a reliability study cited in Reference 10 and has been found to be acceptable through operating experience.
| |
| deleted I.1 Alternatively, a Completion Time If any required Action and associated Completion Time of can be determined Condition G or H is not met, the plant must be brought to a in accordance with MODE in which the overall plant risk is minimized. To achieve this status, the plant must be brought to at least the Risk Informed MODE 3 within 12 hours. Remaining in the Applicability of Completion Time the LCO is acceptable because the plant risk in MODE 3 is Program. similar to or lower than the risk in MODE 4 (Ref. 11) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.5.1-10 Revision 40
| |
| | |
| RCIC System B 3.5.3 BASES (continued) deleted ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable RCIC System. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable RCIC System and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
| |
| A.1 and A.2 If the RCIC System is inoperable during MODE 1, or MODE 2 or 3 with reactor steam dome pressure 150 psig, and the or in accordance HPCI System is immediately verified to be OPERABLE, the RCIC System must be restored to OPERABLE status within 14 days.
| |
| with the Risk In this Condition, loss of the RCIC System will not affect Informed the overall plant capability to provide makeup inventory at Completion Time high reactor pressure since the HPCI System is the only high Program pressure system assumed to function during a loss of coolant accident (LOCA). OPERABILITY of HPCI is therefore immediately verified when the RCIC System is inoperable.
| |
| This may be performed as an administrative check, by examining logs or other information, to determine if HPCI is out of service for maintenance or other reasons. It does not mean it is necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the HPCI System. If the OPERABILITY of the HPCI System cannot be immediately verified, however, Condition B must be entered. For transients and certain abnormal events with no LOCA, RCIC (as opposed to HPCI) is the preferred source of makeup coolant because of its relatively small capacity, which allows easier control of the RPV water level. Therefore, a limited time is allowed to restore the inoperable RCIC to OPERABLE status.
| |
| The 14 day Completion Time is based on a reliability study (Ref. 2) that evaluated the impact on ECCS availability, assuming various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowed outage times (AOTs). Because of similar functions of HPCI and RCIC, the AOTs (i.e., Completion Times) determined for HPCI are also applied to RCIC.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.5.3-3 Revision 22
| |
| | |
| Primary Containment Air Lock B 3.6.1.2 BASES ACTIONS C.1, C.2, and C.3 (continued)
| |
| If the air lock is inoperable for reasons other than those described in Condition A or B, Required Action C.1 requires action to be immediately initiated to evaluate containment overall leakage rates using current air lock leakage test results. An evaluation is acceptable since it is overly conservative to immediately declare the primary containment inoperable if the overall air lock leakage is not within limits. In many instances, primary containment remains OPERABLE, yet only 1 hour (according to LCO 3.6.1.1) would be provided to restore the air lock door to OPERABLE status prior to requiring a plant shutdown. In addition, even with both doors failing the seal test, the overall containment leakage rate can still be within limits.
| |
| Required Action C.2 requires that one door in the primary containment air lock must be verified closed. This action must be completed within the 1 hour Completion Time. This specified time period is consistent with the ACTIONS of LCO 3.6.1.1, which require that primary containment be restored to OPERABLE status within 1 hour.
| |
| or in accordance with the Risk Additionally, the air lock must be restored to OPERABLE Informed status within 24 hours (Required Action C.3). The 24 hour Completion Time is reasonable for restoring an inoperable Completion Time air lock to OPERABLE status considering that at least one Program door is maintained closed in the air lock.
| |
| D.1 and D.2 If the inoperable primary containment air lock cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.6.1.2-6 Revision 0
| |
| | |
| PCIVs B 3.6.1.3 BASES ACTIONS The ACTIONS are modified by Notes 3 and 4. Note 3 ensures (continued) that appropriate remedial actions are taken, if necessary, if the affected system(s) are rendered inoperable by an inoperable PCIV (e.g., an Emergency Core Cooling System subsystem is inoperable due to a failed open test return valve). Note 4 ensures appropriate remedial actions are taken when the primary containment leakage limits are exceeded. Pursuant to LCO 3.0.6, these actions are not required even when the associated LCO is not met.
| |
| Therefore, Notes 3 and 4 are added to require the proper actions be taken.
| |
| A.1 and A.2 With one or more penetration flow paths with one PCIV inoperable, except for MSIV leakage rate not within limit, the affected penetration flow paths must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, a blind flange, and a check valve with flow through the valve secured. For a penetration isolated in accordance with Required Action A.1, the device used to or in accordance isolate the penetration should be the closest available with the Risk valve to the primary containment. The Required Action must Informed be completed within the 4 hour Completion Time (8 hours for Completion Time main steam lines). The Completion Time of 4 hours is Program reasonable considering the time required to isolate the penetration and the relative importance of supporting primary containment OPERABILITY during MODES 1, 2, and 3.
| |
| For main steam lines, an 8 hour Completion Time is allowed.
| |
| The Completion Time of 8 hours for the main steam lines Alternatively, a allows a period of time to restore the MSIVs to OPERABLE status given the fact that MSIV closure will result in Completion Time isolation of the main steam line(s) and a potential for can be determined plant shutdown.
| |
| in accordance with the Risk Informed For affected penetrations that have been isolated in Completion Time accordance with Required Action A.1, the affected penetration flow path(s) must be verified to be isolated on Program.
| |
| a periodic basis. This is necessary to ensure that primary containment penetrations required to be isolated following an accident, and no longer capable of being automatically isolated, will be in the isolation position should an event (continued)
| |
| Quad Cities 1 and 2 B 3.6.1.3-5 Revision 0
| |
| | |
| PCIVs B 3.6.1.3 BASES ACTIONS A.1 and A.2 (continued) occur. This Required Action does not require any testing or device manipulation. Rather, it involves verification that those devices outside primary containment and capable of potentially being mispositioned are in the correct position.
| |
| The Completion Time of "once per 31 days for isolation devices outside primary containment" is appropriate because the devices are operated under administrative controls and the probability of their misalignment is low. For the devices inside primary containment, the time period following specified "prior to entering MODE 2 or 3 from MODE 4, if isolation primary containment was de-inerted while in MODE 4 if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of the inaccessibility of the devices and the existence of other administrative controls ensuring that device misalignment is an unlikely possibility.
| |
| Condition A is modified by a Note indicating that this Condition is only applicable to those penetration flow paths with two or more PCIVs. For penetration flow paths with one PCIV, Condition C provides the appropriate Required Actions.
| |
| Required Action A.2 is modified by two Notes. Note 1 applies to isolation devices located in high radiation areas and allows them to be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignment, once they have been verified to be in the proper position, is low.
| |
| B.1 With one or more penetration flow paths with two or more PCIVs inoperable, except for MSIV leakage rate not within limit, either the inoperable PCIVs must be restored to (continued)
| |
| Quad Cities 1 and 2 B 3.6.1.3-6 Revision 0
| |
| | |
| PCIVs B 3.6.1.3 BASES ACTIONS C.1 and C.2 (continued) penetration (hence, reliability) to act as a penetration isolation boundary and the small pipe diameter of the affected penetrations. In the event the affected penetration flow path is isolated in accordance with Required Action C.1, the affected penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that primary containment penetrations required to be isolated following an accident are isolated.
| |
| This Required Action does not require any testing or valve manipulation. Rather, it involves verification that those devices outside containment and capable of potentially being mispositioned are in the correct position. The Completion Time of once per 31 days is appropriate because the devices are operated under administrative controls and the probability of their misalignment is low.
| |
| following isolation Condition C is modified by a Note indicating that this Condition is only applicable to penetration flow paths with only one PCIV. For penetration flow paths with two or more PCIVs, Conditions A and B provide the appropriate Required Actions. This Note is necessary since this Condition is written specifically to address those penetrations with a single PCIV.
| |
| Required Action C.2 is modified by two Notes. Note 1 applies to isolation devices located in high radiation areas and allows them to be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignment, once they have been verified to be in the proper position, is low.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.6.1.3-8 Revision 0
| |
| | |
| Low Set Relief Valves B 3.6.1.6 BASES APPLICABLE The low set relief mode functions to ensure that the SAFETY ANALYSES containment design basis of no more than two relief valve operating on "subsequent actuations" is met. In other words, multiple simultaneous openings of relief valves (following the initial opening), and the corresponding higher loads, are avoided. The safety analysis demonstrates that the low set relief functions to avoid the induced thrust loads on the relief valve discharge line resulting from "subsequent actuations" of the relief valve during Design Basis Accidents (DBAs). Even though two low set relief valves are specified, only one low set relief valve is required to operate in any DBA analysis.
| |
| The low set relief valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
| |
| LCO Two low set relief valves are required to be OPERABLE to satisfy the assumptions of the safety analyses (Ref. 1).
| |
| The requirements of this LCO are applicable to the mechanical and electrical capability of the low set relief valves to function for controlling the opening and closing of the low set relief valves.
| |
| APPLICABILITY In MODES 1, 2, and 3, an event could cause pressurization of the reactor and opening of relief valves. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the low set relief valves OPERABLE is not required in MODE 4 or 5.
| |
| ACTIONS A.1 With one low set relief valve inoperable, the remaining OPERABLE low set relief valve is adequate to perform the designed function. However, the overall reliability is reduced. The 14 day Completion Time takes into account the redundant capability afforded by the remaining low set relief valve and the low probability of an event occurring during this period in which the remaining low set relief valve capability would be required.
| |
| (continued)
| |
| Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
| |
| Quad Cities 1 and 2 B 3.6.1.6-2 Revision 17
| |
| | |
| Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.7 BASES APPLICABILITY In MODES 4 and 5, the probability and consequences of these (continued) events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining reactor building-to-suppression chamber vacuum breakers OPERABLE is not required in MODE 4 or 5.
| |
| ACTIONS A Note has been added to provide clarification that, for the purpose of this LCO, separate Condition entry is allowed for each reactor building-to-suppression chamber vacuum breaker line.
| |
| A.1 With one or more lines with one vacuum breaker not closed, the leak tight primary containment boundary may be threatened. Therefore, the inoperable vacuum breakers must be restored to OPERABLE status or the open vacuum breaker closed within 7 days. The 7 day Completion Time takes into account the redundancy afforded by the remaining breakers, the fact that the OPERABLE breaker in each of the lines is closed, and the low probability of an event occurring that would require the vacuum breakers to be OPERABLE during this period.
| |
| or in accordance with the Risk Informed Completion Time Program B.1 With one or more lines with two vacuum breakers not closed, primary containment integrity is not maintained. Therefore, one open vacuum breaker must be closed within 1 hour. This Completion Time is consistent with the ACTIONS of LCO 3.6.1.1, "Primary Containment," which requires that primary containment be restored to OPERABLE status within 1 hour.
| |
| C.1 With one line with one or more vacuum breakers inoperable for opening, the leak tight primary containment boundary is intact. The ability to mitigate an event that causes a containment depressurization is threatened, however, if both vacuum breakers in at least one vacuum breaker penetration (continued)
| |
| Quad Cities 1 and 2 B 3.6.1.7-4 Revision 0
| |
| | |
| Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.7 BASES ACTIONS C.1 (continued) are not OPERABLE. Therefore, the inoperable vacuum breaker must be restored to OPERABLE status within 7 days. This is consistent with the Completion Time for Condition A and the fact that the leak tight primary containment boundary is being maintained.
| |
| or in accordance with the Risk Informed D.1 Completion Time Program If one line has one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening and they are not restored within the Completion Time in Condition C, the remaining breakers in the remaining lines can provide the opening function. The plant must be brought to a condition in which the overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 3) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short.
| |
| However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant system.
| |
| Required Action D.1 is modified by a Note that prohibits the application of LCO 3.0.4.a. This Note clarifies the intent of the Required Action by indicating that it is not permissible under LCO 3.0.4.a to enter MODE 3 from MODE 4 with the LCD not met. While remaining in MODE 3 presents an acceptable level of risk, it is not the intent of the Required Action to allow entry into, and continue operation in, MODE 3 from MODE 4 in accordance with LCO 3.0.4.a.
| |
| However, where allowed, a risk assessment may be performed in accordance with LCO 3.0.4.b. Consideration of the results of this risk assessment is required to determine the acceptability of entering MODE 3 from MODE 4 when this LCO is not met.
| |
| (continued) deleted Quad Cities 1 and 2 B 3.6.1.7-5 Revision 40
| |
| | |
| Reactor Building-to-Suppression Chamber Vacuum Breakers B 3.6.1.7 BASES ACTIONS E.1 (continued)
| |
| With two lines with one or more vacuum breakers inoperable for opening, the primary containment boundary is intact.
| |
| However, in the event of a containment depressurization, the function of the vacuum breakers is lost. Therefore, all vacuum breakers in one line must be restored to OPERABLE status within 1 hour. This Completion Time is consistent with the ACTIONS of LCO 3.6.1.1, which requires that primary containment be restored to OPERABLE status within 1 hour.
| |
| or in accordance with the Risk Informed F.1 and F.2 Completion Time Program If any Required Action and associated Completion time of condition A, B, or E can not be met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours and to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| SURVEILLANCE SR 3.6.1.7.1 REQUIREMENTS Each vacuum breaker is verified to be closed to ensure that a potential breach in the primary containment boundary is deleted not present. This Surveillance is performed by observing local or control room indications of vacuum breaker position. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
| |
| Two Notes are added to this SR. The first Note allows reactor-to-suppression chamber vacuum breakers opened in conjunction with the performance of a Surveillance to not be considered as failing this SR. These periods of opening vacuum breakers are controlled by plant procedures and do not represent inoperable vacuum breakers. The second Note is included to clarify that vacuum breakers open due to an actual differential pressure are not considered as failing this SR.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.6.1.7-6 Revision 43
| |
| | |
| Suppression Chamber-to-Drywell Vacuum Breakers B 3.6.1.8 BASES ACTIONS A.1 (continued) would not function as designed during an event that depressurized the drywell), the remaining eight OPERABLE vacuum breakers are capable of providing the vacuum relief function. However, overall system reliability is reduced because additional failures in the remaining vacuum breakers could result in an excessive suppression chamber-to-drywell differential pressure during a DBA. Therefore, with one of the nine required vacuum breakers inoperable, 72 hours is allowed to restore at least one of the inoperable vacuum breakers to OPERABLE status so that plant conditions are consistent with the LCO requirements. The 72 hour deleted Completion Time is considered acceptable due to the low probability of an event in which the remaining vacuum breaker capability would not be adequate.
| |
| Alternatively, a Completion Time B.1 can be determined If a required suppression chamber-to-drywell vacuum breaker in accordance with is inoperable for opening and is not restored to OPERABLE the Risk Informed status within the required Completion Time, the plant must Completion Time be brought to a condition in which the overall plant risk is Program. minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 3) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state. The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
| |
| Required Action B.1 is modified by a Note that prohibits the application of LCO 3.0.4.a. This Note clarifies the intent of the Required Action by indicating that it is not permissible under LCO 3.0.4.a to enter MODE 3 from MODE 4 with the LCO not met. While remaining in MODE 3 presents an acceptable level of risk, it is not the intent of the Required Action to allow entry into, and continue operation in, MODE 3 from MODE 4 in accordance with LCO 3.0.4.a.
| |
| However, where allowed, a risk assessment may be performed in accordance with LCO 3.0.4.b. Consideration of the results (continued)
| |
| Quad Cities 1 and 2 B 3.6.1.8-4 Revision 40
| |
| | |
| RHR Suppression Pool Cooling B 3.6.2.3 BASES APPLICABLE primary containment conditions within design limits. The SAFETY ANALYSES suppression pool temperature is calculated to remain below (continued) the design limit.
| |
| The RHR Suppression Pool Cooling System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
| |
| LCO During a DBA, a minimum of one RHR suppression pool cooling subsystem is required to maintain the primary containment peak pressure and temperature below design limits (Ref. 1).
| |
| To ensure that these requirements are met, two RHR suppression pool cooling subsystems must be OPERABLE with power from two safety related independent power supplies.
| |
| Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. An RHR suppression pool cooling subsystem is deleted OPERABLE when one of the pumps, the heat exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE. Management of gas voids is important to RHR Suppression Pool Cooling System OPERABILITY.
| |
| APPLICABILITY In MODES 1, 2, and 3, a DBA could cause both a release of radioactive material to primary containment and a heatup and pressurization of primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, the RHR Suppression Pool Cooling System is not required to be OPERABLE in MODE 4 or 5.
| |
| ACTIONS A.1 With one RHR suppression pool cooling subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE RHR suppression pool cooling subsystem is adequate to perform the primary containment cooling function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reduced primary containment cooling capability. The 7 day Completion Time is acceptable in light of the redundant RHR suppression pool (continued) or in accordance with the Risk Informed Completion Time Program Quad Cities 1 and 2 B 3.6.2.3-2 Revision 50
| |
| | |
| RHR Drywell Spray B 3.6.2.6 BASES (continued) deleted LCO In the event of a Design Basis Accident (DBA), a minimum of one RHR drywell spray subsystem using one RHR pump is required to adequately scrub the inorganic iodines and particulates from the primary containment atmosphere. To ensure that these requirements are met, two RHR drywell spray subsystems must be OPERABLE with power from two safety related independent power supplies. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. An RHR drywell spray subsystem is OPERABLE when one of the pumps and associated piping, valves, instrumentation, and controls are OPERABLE.
| |
| Management of gas voids is important to RHR drywell spray system OPERABILITY.
| |
| APPLICABILITY In MODES 1, 2, and 3, a DBA could release fission products into the primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES.
| |
| Therefore, maintaining RHR drywell spray subsystems OPERABLE is not required in MODE 4 or 5.
| |
| ACTIONS A.1 With one RHR drywell spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE RHR drywell spray subsystem is adequate to perform the primary containment fission product scrubbing function.
| |
| However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in the loss or in accordance with of the scrubbing capability of the RHR drywell spray system.
| |
| The 7-day Completion Time was chosen in light of the the Risk Informed redundant RHR drywell spray capabilities afforded by the Completion Time OPERABLE subsystem and the low probability of a DBA Program occurring during this period.
| |
| B.1 With both RHR drywell spray subsystems inoperable, at least one subsystem must be restored to OPERABLE status within 8 hours. In this condition, there is a substantial loss of the fission product scrubbing function of the RHR drywell (continued)
| |
| Quad Cities 1 and 2 B 3.6.2.6-2 Revision 68 XX
| |
| | |
| RHRSW System B 3.7.1 BASES ACTIONS A.1 (continued) failure in the OPERABLE subsystem could result in reduced RHRSW capability. The 30 day Completion Time is based on the remaining RHRSW heat removal capability and the low probability of a DBA with concurrent worst case single failure.
| |
| or in accordance with the Risk Informed Completion Time Program B.1 With one RHRSW pump inoperable in each subsystem, the remaining OPERABLE pump in each subsystem can provide adequate heat removal capacity following a design basis LOCA with concurrent worst case single failure. One inoperable pump is required to be restored to OPERABLE status within 7 days. The 7 day Completion Time for restoring one inoperable RHRSW pump to OPERABLE status is based on engineering judgment, considering the level of redundancy provided and low probability of an event occurring requiring RHRSW during this time period.
| |
| C.1 Required Action C.1 is intended to handle the inoperability of one RHRSW subsystem for reasons other than Condition A.
| |
| The Completion Time of 7 days is allowed to restore the RHRSW subsystem to OPERABLE status. With the unit in this condition, the remaining OPERABLE RHRSW subsystem is adequate to perform the RHRSW heat removal function.
| |
| However, the overall reliability is reduced because a single Alternatively, a failure in the OPERABLE RHRSW subsystem could result in loss of RHRSW function. The Completion Time is based on the Completion Time redundant RHRSW capabilities afforded by the OPERABLE can be determined subsystem and the low probability of an event occurring in accordance with requiring RHRSW during this period.
| |
| the Risk Informed Completion Time The Required Action is modified by a Note indicating that the applicable Conditions of LCO 3.4.7, be entered and Program.
| |
| Required Actions taken if the inoperable RHRSW subsystem results in an inoperable RHR shutdown cooling subsystem.
| |
| This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.7.1-4 Revision 0
| |
| | |
| SSMP System B 3.7.9 BASES (continued)
| |
| APPLICABLE The function of the SSMP System is to respond to transient SAFETY ANALYSES events by providing makeup coolant to the reactor. The SSMP System is not an Engineered Safety Feature System and no credit is taken in the safety analyses for SSMP System operation. The system provides a backup to the Unit 1 and 2 RCIC Systems to satisfy the requirements of criteria of 10 CFR 50, Appendix R, Section III.G (Ref. 1). Based on its contribution to the reduction of overall plant risk, the system satisfies Criterion 4 of 10 CFR 50.36 (c)(2)(ii) and is therefore included in the Technical Specifications.
| |
| LCO The OPERABILITY of the SSMP System ensures sufficient reactor water makeup is provided in the event of RPV isolation accompanied by a loss of feedwater flow. The SSMP System has sufficient capacity for maintaining RPV inventory during an isolation event.
| |
| APPLICABILITY The SSMP System is required to be OPERABLE during MODE 1, and MODES 2 and 3 with reactor steam dome pressure 150 psig, since the SSMP System provides a non-Emergency Core Cooling System water source for makeup when the reactor is isolated and pressurized. In MODES 2 and 3 with reactor steam dome pressure 150 psig, and in MODES 4 and 5, the SSMP System is not required to be OPERABLE since the low pressure ECCS injection/spray subsystems can provide sufficient flow to the RPV and since the plant risk associated with fire is also reduced during these MODES.
| |
| or in accordance with the Risk Informed ACTIONS A.1 and A.2 Completion Time Program If the SSMP System is inoperable during MODE 1, or MODE 2 or 3 with reactor steam dome pressure 150 psig, the SSMP System must be restored to OPERABLE status within 14 days.
| |
| In this Condition, loss of the SSMP System will not affect the overall plant capability to provide makeup inventory at high reactor pressure since the RCIC and HPCI System are required to be OPERABLE. The 14 day Completion Time is consistent with the Completion Time for a RCIC System inoperability, because of similar functions of the RCIC and SSMP Systems. The same Completion Time for RCIC is also applied to the SSMP System since the SSMP System and the RCIC System have the same post-fire shutdown functionality goals to provide reactor water makeup (Ref. 3).
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.7.9-2 Revision 0
| |
| | |
| AC SourcesOperating B 3.8.1 BASES ACTIONS A.2 (continued) failure protection may have been lost for the required feature's function; however, function is not lost. The 24 hour Completion Time takes into account the component OPERABILITY of the redundant counterpart to the inoperable required feature. Additionally, the 24 hour Completion Time takes into account the capacity and capability of the remaining AC sources, a reasonable time for repairs, and the low probability of a DBA occurring during this period.
| |
| A.3 With one offsite circuit inoperable, the reliability of the offsite system is degraded, and the potential for a loss of offsite power is increased, with attendant potential for a challenge to the plant safety systems. In this condition, however, the remaining OPERABLE offsite circuit and DGs are adequate to supply electrical power to the onsite Class 1E Distribution System.
| |
| The 7 day Completion Time takes into account the capacity and capability of the remaining AC sources, reasonable time for repairs, and the low probability of a DBA occurring during this period. deleted (continued)
| |
| Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
| |
| Quad Cities 1 and 2 B 3.8.1-9 Revision 63
| |
| | |
| AC SourcesOperating B 3.8.1 BASES ACTIONS B.3.1 and B.3.2 (continued) not to exist on the remaining DG(s), performance of SR 3.8.1.2 suffices to provide assurance of continued OPERABILITY of those DGs.
| |
| In the event the inoperable DG is restored to OPERABLE status prior to completing either B.3.1 or B.3.2, the station corrective action program will continue to evaluate the common cause possibility. This continued evaluation, however, is no longer under the 24 hour constraint imposed while in Condition B.
| |
| If while a DG is inoperable, a new problem with the DG is discovered that would have prevented the DG from performing its specified safety function, a separate entry into Condition B is not required. The new DG problem should be addressed in accordance with the station corrective action program.
| |
| According to Generic Letter 84-15 (Ref. 7), 24 hours is a reasonable time to confirm that the OPERABLE DG(s) are not affected by the same problem as the inoperable DG.
| |
| B.4 In Condition B, the remaining OPERABLE DGs and offsite circuits are adequate to supply electrical power to the onsite Class 1E Distribution System. The 7 day Completion Time takes into account the capacity and capability of the remaining AC sources, a reasonable time for repairs, and the low probability of a DBA occurring during this period. deleted (continued)
| |
| Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
| |
| Quad Cities 1 and 2 B 3.8.1-12 Revision 63
| |
| | |
| AC SourcesOperating B 3.8.1 BASES ACTIONS C.1 and C.2 (continued) transient. In fact, a simultaneous loss of offsite AC sources, a LOCA, and a worst case single failure were postulated as a part of the design basis in the safety analysis. Thus, the 24 hour Completion Time provides a period of time to effect restoration of one of the offsite circuits commensurate with the importance of maintaining an AC electrical power system capable of meeting its design criteria.
| |
| According to Regulatory Guide 1.93 (Ref. 8), with the available offsite AC sources two less than required by the LCO, operation may continue for 24 hours. If two offsite sources are restored within 24 hours, unrestricted operation may continue. If only one required offsite source is restored within 24 hours, power operation continues in accordance with Condition A.
| |
| Alternatively, a Completion Time can be determined in D.1 and D.2 accordance with the Risk Informed Completion Time Program.
| |
| Pursuant to LCO 3.0.6, the Distribution System ACTIONS would not be entered even if all AC sources to it were inoperable, resulting in de-energization. Therefore, the Required Actions of Condition D are modified by a Note to indicate that when Condition D is entered with no AC source to any 4160 V ESS bus (i.e., the bus is de-energized), ACTIONS for LCO 3.8.7, "Distribution SystemsOperating," must be immediately entered. This allows Condition D to provide requirements for the loss of the required offsite circuit and one required DG without regard to whether a division is de-energized. LCO 3.8.7 provides the appropriate restrictions for a de-energized division.
| |
| According to Regulatory Guide 1.93 (Ref. 8), operation may continue in Condition D for a period that should not exceed 12 hours. In Condition D, individual redundancy is lost in both the offsite electrical power system and the onsite AC electrical power system. Since power system redundancy is provided by two diverse sources of power, however, the reliability of the power systems in this Condition may appear higher than that in Condition C (loss of both required offsite circuits). This difference in reliability is offset by the susceptibility of this power system configuration to a single bus or switching failure. The (continued)
| |
| Quad Cities 1 and 2 B 3.8.1-15 Revision 59
| |
| | |
| AC SourcesOperating B 3.8.1 BASES ACTIONS D.1 and D.2 (continued) 12 hour Completion Time takes into account the capacity and capability of the remaining AC sources, reasonable time for repairs, and the low probability of a DBA occurring during this period.
| |
| Alternatively, a Completion Time can be determined in E.1 accordance with the Risk Informed Completion Time Program.
| |
| With two required DGs inoperable, there is no more than one remaining standby AC source. Thus, with an assumed loss of offsite electrical power, sufficient standby AC sources may not be available to power the minimum required ESF functions. Since the offsite electrical power system is the only source of AC power for the majority of ESF equipment at this level of degradation, the risk associated with continued operation for a very short time could be less than that associated with an immediate controlled shutdown. (The immediate shutdown could cause grid instability, which could result in a total loss of AC power.) Since any inadvertent unit generator trip could also result in a total loss of offsite AC power, however, the time allowed for continued operation is severely restricted. The intent here is to avoid the risk associated with an immediate controlled shutdown and to minimize the risk associated with this level of degradation.
| |
| According to Regulatory Guide 1.93 (Ref. 8), with both DGs inoperable, operation may continue for a period that should not exceed 2 hours. The Completion Time assumes complete loss of onsite (DG) AC capability to power the minimum loads needed to respond to analyzed events.
| |
| Alternatively, a Completion Time can be determined in F.1 accordance with the Risk Informed Completion Time Program.
| |
| If the inoperable AC electrical power sources cannot be restored to OPERABLE status within the associated Completion Time, the unit must be brought to a MODE in which the overall plant risk is minimized. To achieve this status, the unit must be brought to at least MODE 3 within 12 hours.
| |
| Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 9) and because the time spent in (continued)
| |
| Quad Cities 1 and 2 B 3.8.1-16 Revision 59
| |
| | |
| DC SourcesOperating B 3.8.4 BASES APPLICABILITY The DC electrical power requirements for MODES 4 and 5 and (continued) other conditions in which the DC electrical power sources are required are addressed in LCO 3.8.5, "DC Sources Shutdown."
| |
| ACTIONS A.1 Condition A represents one 250 VDC electrical power subsystem with a loss of ability to completely respond to an event, and a potential loss of ability to remain energized during normal operation. It is therefore imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for complete loss of 250 VDC power to the affected buses.
| |
| If one 250 VDC electrical power subsystem is inoperable (e.g., inoperable battery, inoperable required battery charger, or inoperable battery charger and associated inoperable battery), the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition. Since a subsequent worst case single failure could, however, result in the loss of minimum necessary 250 VDC electrical subsystems to mitigate a worst case accident, continued power operation should not exceed 72 hours. The Completion Time is based on the capacity and capability of the remaining 250 VDC subsystem.
| |
| Alternatively, a Completion Time can be determined in B.1 and B.2 accordance with the Risk Informed Completion Time Program.
| |
| Condition B, Division 1 or 2 125 VDC battery inoperable as a result of maintenance or testing, represents one division with a loss of ability to completely respond to an event.
| |
| It is therefore imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for complete loss of DC power to the affected division.
| |
| Operation in this Condition is needed during the operating cycle to ensure the battery is maintained OPERABLE.
| |
| Condition B is modified by a Note indicating that the Condition is only applicable when the opposite unit is in MODE 1, 2, or 3.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.8.4-5 Revision 0
| |
| | |
| DC SourcesOperating B 3.8.4 BASES ACTIONS B.1 and B.2 (continued)
| |
| If one of the 125 VDC batteries is inoperable, the remaining 125 VDC electrical power subsystem has the capacity to support a safe shutdown of one unit and to mitigate an accident condition in the other unit. Since a subsequent worst case single failure could, however, result in the loss of minimum necessary DC electrical subsystems to mitigate a worst case accident, continued power operation is limited.
| |
| Required Action B.2 limits the time the unit can operate in this condition to 7 cumulative days per operating cycle, for any one battery. Therefore, each 125 VDC battery can be removed from service to perform maintenance or testing as Alternatively, a long as the cumulative time is not exceeded for that Completion Time can be battery. In addition, Required Action B.1 requires the associated OPERABLE alternate 125 VDC electrical power determined in subsystem to be placed in service. An OPERABLE alternate accordance with the Risk 125 VDC electrical power subsystem consists of the alternate Informed Completion 125 VDC battery and one full capacity battery charger. For Time Program. the alternate 125 VDC battery to be considered OPERABLE, all SR requirements associated with the alternate 125 VDC battery must be met. (The full capacity battery charger is the same battery charger (normal or spare) associated with the normal 125 VDC electrical power subsystem.) Therefore, placement of the OPERABLE alternate 125 VDC electrical power subsystem in service will help ensure that the design basis can be met. However, the design configuration of the alternate battery is susceptible to single failure and hence, is not as reliable as the normal battery. Therefore, only a limited time of operation is allowed in this condition.
| |
| The 72 hour Completion Time to place the associated OPERABLE alternate 125 VDC electrical power subsystem in service provides sufficient time to safely remove the Division 1 or 2 125 VDC electrical power subsystem from service and place the alternate supply in service. The 7 day cumulative Completion Time is based on the capacity and capability of the remaining DC Sources, including the enhanced capability afforded by the capability of the alternate 125 VDC electrical power subsystem to supply the required loads.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.8.4-6 Revision 0
| |
| | |
| DC SourcesOperating B 3.8.4 BASES ACTIONS C.1 and C.2 (continued)
| |
| Condition C, Division 1 or 2 125 VDC battery inoperable due to the need to replace the battery as determined by maintenance or testing, represents one division with a loss of ability to completely respond to an event. It is therefore imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for complete loss of DC power to the affected division. Operation in this Condition may be needed during the operating cycle to completely replace a battery to maintain the Division 1 or 2 VDC subsystem OPERABLE for the remainder of the cycle.
| |
| Condition C is modified by a Note indicating that the Condition is only applicable when the opposite unit is in MODE 1, 2, or 3.
| |
| If one of the 125 VDC batteries is inoperable, the remaining 125 VDC electrical power subsystem has the capacity to support a safe shutdown of one unit and to mitigate an accident condition in the other unit. Since a subsequent worst case single failure could, however, result in the loss of minimum necessary DC electrical subsystems to mitigate a worst case accident, continued power operation is limited.
| |
| Required Action C.2 limits the time the unit can operate in this condition to 7 days. Therefore, each 125 VDC battery can be removed from service to completely replace a battery.
| |
| In addition, Required Action C.1 requires the associated OPERABLE alternate 125 VDC electrical power subsystem to be or in accordance placed in service. An OPERABLE alternate 125 VDC electrical power subsystem consists of the alternate 125 VDC battery with the Risk and one full capacity battery charger. For the alternate Informed 125 VDC battery to be considered OPERABLE, all SR Completion Time requirements associated with the alternate 125 VDC battery Program must be met. (The full capacity battery charger is the same battery charger (normal or spare) associated with the normal 125 VDC electrical power subsystem.) Therefore, placement of the OPERABLE alternate 125 VDC electrical power subsystem in service will help ensure that the design basis can be met. However, the design configuration of the alternate battery is susceptible to single failure and hence, is not as reliable as the normal battery. Therefore, only a limited time of operation is allowed in this condition.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.8.4-7 Revision 0
| |
| | |
| DC SourcesOperating B 3.8.4 BASES ACTIONS D.1 and D.2 (continued) requires the OPERABLE alternate 125 VDC electrical power subsystem to be placed in service in 72 hours. The 72 hour Completion Time to place associated OPERABLE alternate or in accordance 125 VDC electrical power subsystem in service provides with the Risk sufficient time to safely remove the Division 1 or 2 125 VDC Informed electrical power subsystem from service and place the Completion Time alternate supply in service. An OPERABLE alternate 125 VDC Program electrical power subsystem consists of the alternate 125 VDC battery and one full capacity battery charger. For the alternate 125 VDC battery to be considered OPERABLE, all SR requirements associated with the alternate 125 VDC battery must be met. (The full capacity battery charger is the same battery charger (normal or spare) associated with the normal 125 VDC electrical power subsystem.) Upon completing this Required Action continuous operation is allowed, since if the opposite unit associated OPERABLE alternate 125 VDC electrical power subsystem is placed in service supplying the unit Division 2 loads, the design configuration will not be susceptible to single failure and hence, the reliability is consistent with the normal circuit.
| |
| E.1 With the opposite unit Division 2 125 VDC electrical power system inoperable, certain redundant Division 2 features (e.g., Standby Gas Treatment System) will not function if a design basis event were to occur. With a standby gas treatment subsystem inoperable, LCO 3.6.4.3, "Standby Gas Treatment System" requires restoration of the inoperable SGT subsystem to OPERABLE status in 7 days. Therefore, a 7 day Completion Time is provided to restore the opposite unit Division 2 125 VDC electrical power subsystem to OPERABLE status. The 7 day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant system(s) and the low probability of a DBA occurring during this time period.
| |
| (continued)
| |
| Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
| |
| Quad Cities 1 and 2 B 3.8.4-9 Revision 0
| |
| | |
| Distribution SystemsOperating B 3.8.7 BASES (continued)
| |
| APPLICABILITY The electrical power distribution subsystems are required to be OPERABLE in MODES 1, 2, and 3 to ensure that:
| |
| : a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and
| |
| : b. Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.
| |
| Electrical power distribution subsystem requirements for MODES 4 and 5 and other conditions in which AC and DC electrical power distribution subsystems are required are covered in the Bases for LCO 3.8.8, "Distribution Systems-Shutdown."
| |
| ACTIONS A.1 With one or more required AC buses, motor control centers, or distribution panels inoperable and a loss of function has not yet occurred, the remaining AC electrical power distribution subsystems are capable of supporting the minimum safety functions necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming no single failure. The overall reliability is reduced, however, because a single failure in the remaining electrical power distribution subsystems could result in the minimum required ESF functions not being supported.
| |
| Therefore, the required AC buses, motor control centers, and distribution panels must be restored to OPERABLE status within 8 hours.
| |
| The Condition A worst scenario is one division without AC Alternatively, a power (i.e., no offsite power to the division and the Completion Time can be associated DG inoperable). In this situation, the unit is more vulnerable to a complete loss of AC power. It is, determined in therefore, imperative that the unit operators' attention be accordance with the Risk focused on minimizing the potential for loss of power to the Informed Completion remaining division by stabilizing the unit and restoring Time Program. power to the affected division. The 8 hour time limit before requiring a unit shutdown in this Condition is acceptable because of:
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.8.7-5 Revision 0
| |
| | |
| Distribution SystemsOperating B 3.8.7 BASES ACTIONS B.1 (continued)
| |
| With one or more DC buses inoperable and a loss of safety function has not yet occurred, the remaining DC electrical power distribution subsystem is capable of supporting the minimum safety functions necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming no single failure. The overall reliability is reduced, however, because a single failure in the remaining DC electrical power distribution subsystem could result in the minimum required ESF functions not being supported.
| |
| Therefore, the required DC electrical power distribution subsystem(s) must be restored to OPERABLE status within 2 hours by powering the bus from the associated battery or charger.
| |
| or in accordance with the Risk Informed Completion Condition B worst scenario is one subsystem without adequate DC power, potentially with both the battery significantly Time Program degraded and the associated charger nonfunctioning. In this situation the plant is significantly more vulnerable to a complete loss of all DC power. It is, therefore, imperative that the operator's attention focus on stabilizing the plant, minimizing the potential for loss of power to the remaining subsystem, and restoring power to the affected subsystem.
| |
| This 2 hour limit is more conservative than Completion Times allowed for the majority of components that would be without power. Taking exception to LCO 3.0.2 for components without adequate DC power, which would have Required Action Completion Times shorter than 2 hours, is acceptable because of:
| |
| : a. The potential for decreased safety when requiring a change in plant conditions (i.e., requiring a shutdown) while not allowing stable operations to continue;
| |
| : b. The potential for decreased safety when requiring entry into numerous applicable Conditions and Required Actions for components without DC power, while not providing sufficient time for the operators to perform the necessary evaluations and actions for restoring power to the affected division; (continued)
| |
| Quad Cities 1 and 2 B 3.8.7-7 Revision 0
| |
| | |
| Distribution SystemsOperating B 3.8.7 BASES ACTIONS C.1 (continued) systems are powered only from Unit 1, an inoperable Unit 1 AC electrical power distribution subsystem could result in a loss of the CREV System and Control Room Emergency Ventilation AC System functions (for both units).
| |
| With a standby gas treatment (SGT) subsystem inoperable, LCO 3.6.4.3 requires restoration of the inoperable SGT subsystem to OPERABLE status in 7 days. Similarly, with the CREV System inoperable, LCO 3.7.4 requires restoration of the inoperable CREV System to OPERABLE status within 7 days.
| |
| Alternatively, a With the Control Room Emergency Ventilation AC System inoperable, LCO 3.7.5 requires restoration of the inoperable Completion Time can be Control Room Emergency Ventilation AC System to OPERABLE determined in status in 30 days. Therefore, a 7 day Completion Time is accordance with the Risk provided to restore the required opposite unit AC and DC Informed Completion electrical power subsystems to OPERABLE status. The 7 day Time Program. Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant system(s) and the low probability of a DBA occurring during this time period.
| |
| The Required Action is modified by a Note indicating that the applicable Conditions of LCO 3.8.1 be entered and Required Actions taken if the inoperable opposite unit AC electrical power distribution subsystem results in an inoperable required offsite circuit. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.
| |
| D.1 deleted If the inoperable distribution subsystem cannot be restored to OPERABLE status within the associated Completion Time, the unit must be brought to a MODE in which the overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours.
| |
| Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 4) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.
| |
| (continued)
| |
| Quad Cities 1 and 2 B 3.8.7-9 Revision 40
| |
| | |
| ATTACHMENT 4 Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Cross-Reference of TSTF-505 and QCNPS Technical Specifications
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 1 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Completion Times 1.3 1.3 Example 1.3-8 1.3-8 1.3-8 TSTF-505 changes are incorporated.
| |
| Standby Liquid Control (SLC) 3.1.7 3.1.7 System One SLC subsystem inoperable QCNPS TS 3.1.7 does not include NUREG-1433
| |
| [for reasons other than Condition optional Condition 1 or the associated Condition 2 A]. wording.
| |
| 3.1.7.B.1 3.1.7.A.1 Yes Restore SLC subsystem to TSTF-505 changes are incorporated.
| |
| OPERABLE status.
| |
| Although Conditions and Required Actions numbering between the NUREG-1433 and QCNPS Technical Specifications is different, this is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 2 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Reactor Protection System 3.3.1.1 3.3.1.1 (RPS) Instrumentation One or more required channels TSTF-505 changes are incorporated.
| |
| inoperable. However, under certain circumstances, for one or more Functions with one or more required channels Place channel in trip. 3.3.1.1.A.1 3.3.1.1.A.1 Yes inoperable, a loss of function may occur. Therefore, a note is added which prohibits applying a RICT when RPS trip capability is not maintained.
| |
| Place associated trip system in 3.3.1.1.A.2 3.3.1.1.A.2 Yes trip.
| |
| One or more Functions with one or TSTF-505 changes are incorporated.
| |
| more required channels However, under certain circumstances, for one or inoperable in both trip systems. more Functions with one or more required channels inoperable, a loss of function may occur. Therefore, 3.3.1.1.B.1 3.3.1.1.B.1 Yes a note is added which prohibits applying a RICT Place channel in one trip system in trip. when RPS trip capability is not maintained.
| |
| Place one trip system in trip. 3.3.1.1.B.2 3.3.1.1.B.2 Yes
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 3 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Source Range Monitor (SRM) 3.3.1.2 Instrumentation One or more required SRMs inoperable in MODE 2 with intermediate range monitors (IRMs) on Range 2 or below.
| |
| Restore required SRMs to 3.3.1.2.A.1 3.3.2.1.A.1 No SRMs not modeled in the QCNPS PRA. Therefore, OPERABLE status. the TSTF-505 change is not incorporated into the QCNPS TS.
| |
| Feedwater and Main Turbine 3.3.2.2 3.3.2.2 QCNPS TS titled "Feedwater System and Main High Water Level Trip Turbine High Water Level Trip Instrumentation."
| |
| Instrumentation One feedwater and main turbine Corresponding Condition A wording is "One or more high water level trip channel Feedwater System and one main turbine high water inoperable. level trip channels inoperable." And includes a note that Required Action A.1 is not applicable if the inoperable channel is the result of an inoperable Place channel in trip. 3.3.2.2.A.1 3.3.2.2.A.1 Yes feedwater pump breaker. These wording differences are administrative in nature and does not impact the applicability of TSTF-505 to this TS.
| |
| TSTF-505 changes are incorporated.
| |
| However, under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 4 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Two or more feedwater and main turbine high water level trip channels inoperable.
| |
| Restore feedwater and main 3.3.2.2.B.1 3.3.2.2.B.1 No QCNPS TS do not contain this TS, as more than 1 turbine high water level trip channel inoperable is covered by QCNPS capability. Condition A wording. Therefore, no change is proposed to the QCNPS TS.
| |
| End of Cycle Recirculation 3.3.4.1 ------------------------ QCNPS TS do not contain this TS. Therefore, a Pump Trip (EOC-RPT) change is not proposed to the QCNPS TS.
| |
| Instrumentation Anticipated Transient Without 3.3.4.2 3.3.4.1 QCNPS TS numbering is different than Scram Recirculation Pump Trip NUREG-1433.
| |
| (ATWS-RPT) Instrumentation Although the Technical Specifications numbering between the NUREG-1433 and QCNPS is different, this is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
| |
| One or more channels inoperable.
| |
| Restore channel to OPERABLE 3.3.4.2.A.1 3.3.4.1.A.1 Yes TSTF-505 changes are incorporated.
| |
| status. However, under certain circumstances, with more than one channel inoperable, a loss of function may Place channel in trip. 3.3.4.2.A.2 3.3.4.1.A.2 Yes occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 5 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Emergency Core Cooling 3.3.5.1 3.3.5.1 System (ECCS) Instrumentation As required by Required Action TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table However, under certain circumstances, with more 3.3.5.1-1. than one channel inoperable, a loss of function may 3.3.5.1.B.3 3.3.5.1.B.3 Yes occur. Therefore, a Note is added to the Completion Place channel in trip. Time which prohibits applying a RICT when trip capability is not maintained.
| |
| As required by Required Action TS 3.3.5.1 - ECCS Instrumentation Condition C A.1 and referenced in Table Required Action C.1. NUREG-1433 mark-up 3.3.5.1-1. contains a note which limits the applicability to functions 1c, 2c, 2d, and 2f. The QCNPS TS note limits the applicability to functions 1c, 1e, 2c, 2e, 2g, 2h, 2f, and 2k. These functions meet all the criteria 3.3.5.1.C.2. 3.3.5.1.C.2 Yes established by the TSTF and therefore are included Restore channel to OPERABLE in the QCNPS RICT TS pages mark-ups. This status. submittal does not change the instrumentation tables.
| |
| TSTF-505 changes are incorporated.
| |
| However, under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 6 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| As required by Required Action TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table However, under certain circumstances, with more 3.3.5.1-1. than one channel inoperable, a loss of function may 3.3.5.1.D.2.1 3.3.5.1.D.2.1 Yes occur. Therefore, a Note is added to the Completion Place channel in trip. Time for D.2.1 which prohibits applying a RICT when trip capability is not maintained.
| |
| As required by Required Action TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table However, under certain circumstances, with more 3.3.5.1-1. than one channel inoperable, a loss of function may 3.3.5.1.E.2 3.3.5.1.E.2 Yes occur. Therefore, a Note is added to the Completion Restore channel to OPERABLE Time which prohibits applying a RICT when trip status. capability is not maintained.
| |
| As required by Required Action TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table However, under certain circumstances, with more 3.3.5.1-1. than one channel inoperable, a loss of function may 3.3.5.1.F.2 3.3.5.1.F.2 Yes occur. Therefore, a Note is added to the Completion Place channel in trip. Time which prohibits applying a RICT when trip capability is not maintained.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 7 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| As required by Required Action TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table However, under certain circumstances, with more 3.3.5.1-1. than one channel inoperable, a loss of function may 3.3.5.1.G.2 3.3.5.1.G.2 Yes occur. Therefore, a Note is added to the Completion Restore channel to OPERABLE Time which prohibits applying a RICT when trip status. capability is not maintained.
| |
| Reactor Core Isolation Cooling 3.3.5.2 3.3.5.3 QCNPS TS numbering is different than (RCIC) System Instrumentation NUREG-1433.
| |
| Although the Technical Specifications numbering between the NUREG-1433 and QCNPS is different, this is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
| |
| As required by Required Action QCNPS TS cross-references Table 3.3.5.3-1, A.1 and referenced in Table consistent with the overall TS section numbering 3.3.5.2-1. difference discussed above.
| |
| 3.3.5.2.B.2 3.3.5.3.B.2 Yes TSTF-505 changes are incorporated.
| |
| Place channel in trip. However, under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 8 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| As required by Required Action QCNPS TS cross-references Table 3.3.5.3-1, A.1 and referenced in Table consistent with the overall TS section numbering 3.3.5.2-1. difference discussed above.
| |
| 3.3.5.2.D.2.1 3.3.5.3.D.2.1 Yes TSTF-505 changes are incorporated.
| |
| Place channel in trip. However, under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
| |
| Primary Containment Isolation 3.3.6.1 3.3.6.1 Instrumentation One or more required channels TS 3.3.6.1 - Primary Containment Isolation inoperable. Instrumentation Condition A Required Action A.1 Completion Time in the NUREG-1433 mark-up contains a note which limits the applicability to Place channel in trip. 3.3.6.1.A.1 3.3.6.1.A.1 Yes functions 2.a, 2.b, 6.b, 7.a and 7.b. The QCNPS TS limits the applicability to functions 1.a, 2.b, 3.d, 5.b, (Functions 1.a, 2.a, and 6.b. These functions meet all the criteria of 2.b, 3.d, 5.b, and 6.b; established by the TSTF and therefore are included Functions other than in the QCNPS RICT TS pages mark-ups. This Functions 1.a, 2.a, submittal does not change the instrumentation tables.
| |
| 2.b, 3.d, 5.b, and 6.b)
| |
| TSTF-505 changes are incorporated.
| |
| However, under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 9 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Low-Low-Set (LLS) 3.3.6.3 3.3.6.3 The QCNPS TS 3.3.6.3 is titled "Relief Valve Instrumentation Instrumentation."
| |
| One LLS valve inoperable due to 3.3.6.3.A.1 3.3.6.3.A.1 Yes The QCNPS TS 3.3.6.3 Condition A reads "One relief inoperable channel(s). valve inoperable due to inoperable channel" and has the same Required Action to restore the channel to operable status as NUREG-1433.
| |
| Restore channel(s) to OPERABLE status. However, under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
| |
| Loss of Power (LOP) 3.3.8.1 3.3.8.1 Instrumentation One or more channels inoperable. TSTF-505 changes are incorporated.
| |
| However, under certain circumstances, with more Place channel in trip. 3.3.8.1.A.1 3.3.8.1.A.1 Yes than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 10 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Safety/Relief Valves 3.4.3 3.4.3 QCNPS TS title is "Safety and Relief Valves."
| |
| QCNPS has 9 SVs and 5 RVs covered by this TS.
| |
| One [or two] [required] S/RV[s] The corresponding QCNPS Condition wording is inoperable. "One relief valve inoperable." With the Required Action to "Restore relief valve to OPERABLE status."
| |
| These wording differences are administrative in Restore the [required] S/RV[s] to 3.4.3.A.1 3.4.3.A.1 Yes nature and do not impact the applicability of OPERABLE status. TSTF-505 to this TS.
| |
| TSTF-505 changes are incorporated.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 11 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| ECCS - Operating 3.5.1 3.5.1 QCNPS Condition wording and numbering is different than NUREG-1433.
| |
| The QCNPS ECCS network consists of the High Pressure Coolant Injection (HPCI) System, the Core Spray (CS) System, the Low Pressure Coolant Injection (LPCI) mode of the Residual Heat Removal (RHR) System, and the Automatic Depressurization System (ADS).
| |
| One low pressure ECCS injection QCNPS LPCI has 2 pumps per subsystem and CS spray subsystem inoperable. has 1 pump per subsystem.
| |
| Condition numbering is different from NUREG-1433.
| |
| OR Condition wording is different in that QCNPS uses "One LPCI subsystem inoperable for reasons other than Condition A" where Condition A is 1 LPCI pump One LPCI pump in both LPCI inoperable. Additionally, the 2nd half of the OR subsystems inoperable. statement is "One Core Spray subsystem inoperable." Required Action wording is the same as NUREG-1433.
| |
| Restore low pressure ECCS injection/spray subsystem to 3.5.1.A.1 3.5.1.B.1 Yes TSTF-505 changes are incorporated.
| |
| OPERABLE status.
| |
| Although Conditions and Required Actions numbering between the NUREG-1433 and QCNPS Technical Specifications is different, this is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 12 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| One LPCI pump in each subsystem inoperable.
| |
| Restore one LPCI pump to ------------------ 3.5.1.C.1 Yes Functionally equivalent to the 2nd half of the OPERABLE status. NUREG-1433 Condition A OR statement and thus meets the same intent as NUREG-1433 markup in TSTF-505.
| |
| Two LPCI subsystems inoperable In this condition, both LPCI subsystems are lost, but for reasons other than the remaining OPERABLE CS subsystems provide Condition C. adequate core cooling during a LOCA.
| |
| Not included in NUREG-1433 markup in TSTF-505,
| |
| ------------------ 3.5.1.E.1 Yes but desired for inclusion in the QCNPS RICT Restore one LPCI subsystem to OPERABLE status. Program. This is acceptable because the Traveler states that there may be plant-specific TS, which meet the 18 criteria described in the Traveler, that may be included in the RICT Program scope.
| |
| HPCI System inoperable. QCNPS TS numbering is different than NUREG-1433.
| |
| Restore HPCI system to 3.5.1.C.1 3.5.1.G.2 Yes OPERABLE status. TSTF-505 changes are incorporated.
| |
| Although Conditions and Required Actions numbering between the NUREG-1433 and QCNPS Technical Specifications is different, this is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 13 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| HPCI System inoperable AND Condition A entered.
| |
| Restore HPCI system to 3.5.1.D.1 --------------------------- No The QCNPS TS do not contain this TS, therefore, OPERABLE status. TSTF-505 changes are not incorporated.
| |
| Restore low pressure ECCS 3.5.1.D.2 -------------------------- No The QCNPS TS do not contain this TS, therefore, injection/spray subsystem(s) to TSTF-505 changes are not incorporated.
| |
| OPERABLE status.
| |
| One ADS valve inoperable. QCNPS TS numbering is different than NUREG-1433.
| |
| Restore ADS valve to OPERABLE 3.5.1.E.1 3.5.1.H.1 Yes status. TSTF-505 changes are incorporated.
| |
| Although Conditions and Required Actions numbering between the NUREG-1433 and QCNPS Technical Specifications is different, this is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 14 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| One ADS valve inoperable AND Condition A entered.
| |
| Restore ADS valve to OPERABLE 3.5.1.F.1 --------------------------- No The QCNPS TS do not contain this TS, therefore, status. TSTF-505 changes are not incorporated.
| |
| Restore low pressure ECCS 3.5.1.F.2 --------------------------- No The QCNPS TS do not contain this TS; therefore, injection/spray subsystem(s) to TSTF-505 changes are not incorporated.
| |
| OPERABLE status.
| |
| RCIC [Reactor Core Isolation 3.5.3 3.5.3 Cooling] System RCIC system inoperable.
| |
| Restore RCIC System to 3.5.3.A.2 3.5.3.A.2 Yes TSTF-505 changes are incorporated.
| |
| OPERABLE status.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 15 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Primary Containment Air Lock 3.6.1.2 3.6.1.2 Primary containment air lock inoperable for reasons other than Condition A or B.
| |
| Restore air lock to OPERABLE 3.6.1.2.C.3 3.6.1.2.C.3 Yes TSTF-505 changes are incorporated.
| |
| status.
| |
| However, under certain circumstances, with more than one primary containment airlock inoperable, excessive leakage or a loss of function may occur.
| |
| Therefore, a Note is added to the Completion Time which prohibits applying a RICT when leakage exceeds limits or there is a loss of function.
| |
| Primary Containment Isolation 3.6.1.3 3.6.1.3 Valves (PCIVs)
| |
| One or more penetration flow paths with one PCIV inoperable for reasons other than Condition[s]
| |
| D [and E].
| |
| Isolate the affected penetration 3.6.1.3.A.1 3.6.1.3.A.1 Yes TSTF-505 changes are incorporated.
| |
| flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 16 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| [One or more penetration flow 3.6.1.3.E.1 --------------------------- No The QCNPS TS do not contain this TS, therefore, paths with one or more TSTF-505 changes are not incorporated.
| |
| containment purge valves not within purge valve leakage limits.]
| |
| Low Set Relief Valves 3.6.1.6 One low set relief valve inoperable.
| |
| Restore low set relief valve to ------------------ 3.6.1.6.A Yes Not included in NUREG-1433 markup in TSTF-505, OPERABLE status. but desired for inclusion in the QCNPS RICT Program. This is acceptable because the Traveler states that there may be plant-specific TS, which meet the 18 criteria described in the Traveler, that may be included in the RICT Program scope.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 17 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Reactor Building-to- 3.6.1.7 3.6.1.7 Suppression Chamber Vacuum Breakers One or more lines with one reactor building-to-suppression chamber vacuum breaker not closed.
| |
| Close the open vacuum breaker. ------------------ 3.6.1.7.A.1 Yes Not included in NUREG-1433 markup in TSTF-505, but desired for inclusion in the QCNPS RICT Program. This is acceptable because the Traveler states that there may be plant-specific TS, which meet the 18 criteria described in the Traveler, that may be included in the RICT Program scope.
| |
| One line with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening.
| |
| Restore the vacuum breaker(s) to 3.6.1.7.C.1 3.6.1.7.C.1 Yes TSTF-505 change is incorporated.
| |
| OPERABLE status.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 18 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Two lines with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening.
| |
| Restore all vacuum breakers in 3.6.1.7.D.1 3.6.1.7.E.1 Yes TSTF-505 change is incorporated.
| |
| one line to OPERABLE status. Although Conditions and Required Actions numbering between the NUREG-1433 and QCNPS Technical Specifications is different, this is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
| |
| Suppression Chamber-to- 3.6.1.8 3.6.1.8 Drywell Vacuum Breakers One required suppression chamber-to-drywell vacuum breaker inoperable for opening.
| |
| Restore one vacuum breaker to 3.6.1.8.A.1 3.6.1.8.A.1 Yes TSTF-505 changes are incorporated.
| |
| OPERABLE status.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 19 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Residual Heat Removal (RHR) 3.6.2.3 3.6.2.3 Suppression Pool Cooling One RHR suppression pool cooling subsystem inoperable.
| |
| Restore RHR suppression pool 3.6.2.3.A.1 3.6.2.3.A.1 Yes TSTF-505 changes are incorporated.
| |
| cooling subsystem to OPERABLE status.
| |
| Residual Heat Removal (RHR) 3.6.2.4 3.6.2.4 Suppression Pool Spray One RHR suppression pool spray subsystem inoperable.
| |
| Restore RHR suppression pool 3.6.2.4.A.1 3.6.2.4.A.1 No Suppression pool spray function not modeled for spray subsystem to OPERABLE RHR, only modeled for failed flow to RHR injection status. valves. Therefore, the TSTF-505 change is not adopted for the QCNPS TS.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 20 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Residual Heat Removal (RHR) ------------------ 3.6.2.6 Drywell Spray One RHR drywell spray The QCNPS LOCA radiological dose analysis credits subsystem inoperable. the RHR drywell spray system for scrubbing radionuclides from the drywell air space.
| |
| Restore RHR drywell spray ------------------ 3.6.2.6.A.1 Yes subsystem to OPERABLE status. Not included in NUREG-1433 markup in TSTF-505, but desired for inclusion in the QCNPS RICT Program. This is acceptable because the Traveler states that there may be plant-specific TS, which meet the 18 criteria described in the Traveler, that may be included in the RICT Program scope.
| |
| Drywell Cooling System Fans 3.6.3.1 ------------------------ No The QCNPS TS do not contain this TS, therefore, TSTF-505 changes are not incorporated.
| |
| Residual Heat Removal Service 3.7.1 3.7.1 Water (RHRSW) System One RHRSW pump in each subsystem inoperable.
| |
| Restore RHRSW pump to 3.7.1.B.1 3.7.1.B.1 Yes TSTF-505 changes are incorporated.
| |
| OPERABLE status.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 21 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| One RHRSW subsystem inoperable for reasons other than Condition A.
| |
| Restore RHRSW subsystem to 3.7.1.C.1 3.7.1.C.1 Yes TSTF-505 changes are incorporated.
| |
| OPERABLE status.
| |
| [Plant Service Water (PSW)] 3.7.2 .. No The QCNPS TS do not contain this TS, therefore, System and [Ultimate Heat Sink TSTF-505 changes are not incorporated.
| |
| (UHS)]
| |
| Main Turbine Bypass System 3.7.7 3.7.7
| |
| [Requirements of the LCO not met QCNPS TS 3.7.6, Condition A, Required Action A.1 or Main Turbine Bypass System requires satisfying the requirements of the LCO, i.e.,
| |
| inoperable]. applying the operating limits, and does not have a restore action.
| |
| [Satisfy the requirements of the 3.7.7.A.1 3.7.7.A.1 No LCO or restore Main Turbine The assumptions of the base-case transient analyses Bypass System to Operable are not satisfied when the Main Turbine Bypass status.] System is not operable; the required action is to implement the alternate MCPR and LHGR limits that have been determined based on an inoperable Main Turbine Bypass system. QCNPS determined it was not appropriate to apply RICT in this situation, therefore, TSTF-505 changes are not incorporated.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 22 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Safe Shutdown Makeup Pump 3.7.9 (SSMP) System SSMP System inoperable.
| |
| Restore SSMP System to 3.7.9.A.1 Yes This TS is unique to QCNPS and supports Fire OPERABLE status. Protection safe shutdown requirements. Therefore, it is included in NUREG-1433 markup in TSTF-505.
| |
| AC Sources - Operating 3.8.1 3.8.1 One [required] offsite circuit inoperable.
| |
| Restore [required] offsite circuit to 3.8.1.A.3 3.8.1.A.3 Yes TSTF-505 changes are incorporated.
| |
| OPERABLE status.
| |
| One [required] DG inoperable.
| |
| Restore [required] DG to 3.8.1.B.4 3.8.1.B.4 Yes TSTF-505 changes are incorporated.
| |
| OPERABLE status.
| |
| Two [required] offsite circuits inoperable.
| |
| Restore one [required] offsite 3.8.1.C.2 3.8.1.C.2 Yes TSTF-505 changes are incorporated.
| |
| circuit to OPERABLE status.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 23 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| One [required] offsite circuit inoperable.
| |
| AND One [required] DG inoperable.
| |
| Restore [required] offsite circuit to 3.8.1.D.1 3.8.1.D.1 Yes TSTF-505 changes are incorporated.
| |
| OPERABLE status.
| |
| Restore [required] DG to 3.8.1.D.2 3.8.1.D.2 Yes TSTF-505 changes are incorporated.
| |
| OPERABLE status.
| |
| [One [required] [automatic load 3.8.1.F.1 ------------------------- No The QCNPS TS do not contain this TS, therefore, sequencer] inoperable. TSTF-505 changes are not incorporated.
| |
| DC Sources - Operating 3.8.4 3.8.4 QCNPS TS is organized to take advantage of the availability of the opposite unit's 125 VDC subsystem for operability of certain shared safety-related equipment.
| |
| One [or two] battery charger[s on one division] inoperable.
| |
| Restore battery charger[s] to 3.8.4.A.3 -------------------------- No The QCNPS TS do not contain this TS, therefore, OPERABLE status. TSTF-505 changes are not incorporated.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 24 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| One [or two] batter[y] [ies] on one 3.8.4.B.1 --------------------------- No The QCNPS TS do not contain this TS, therefore, division] inoperable. TSTF-505 changes are not incorporated.
| |
| Restore batter[y][ies] to OPERABLE status.
| |
| One DC electrical power subsystems inoperable for reasons other than Condition A [or B].
| |
| The QCNPS TS do not contain this TS, therefore, Restore DC electrical power 3.8.4.C.1 ------------------------- No TSTF-505 changes are not incorporated.
| |
| subsystem to OPERABLE status.
| |
| One 250 VDC electrical power subsystem inoperable.
| |
| Restore the 250 VDC electrical ---------------- 3.8.4.A.1 Yes Not included in NUREG-1433 markup in TSTF-505, power subsystem to OPERABLE but desired for inclusion in the QCNPS RICT status. Program. This is acceptable because the Traveler states that there may be plant-specific TS, which meet the 18 criteria described in the Traveler, that may be included in the RICT Program scope.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 25 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Division 1 or 2 125 VDC battery inoperable as a result of maintenance or testing.
| |
| Restore Division 1 or 2 125 VDC ---------------- 3.8.4.B.2 Yes Not included in NUREG-1433 markup in TSTF-505, battery to OPERABLE status. but desired for inclusion in the QCNPS RICT Program. This is acceptable because the Traveler states that there may be plant-specific TS, which meet the 18 criteria described in the Traveler, that may be included in the RICT Program scope.
| |
| Division 1 or 2 125 VDC battery inoperable, due to the need to replace the battery, as determined by maintenance or testing.
| |
| Restore Division 1 or 2 125 VDC ---------------- 3.8.4.C.2 Yes Not included in NUREG-1433 markup in TSTF-505, battery to OPERABLE status. but desired for inclusion in the QCNPS RICT Program. This is acceptable because the Traveler states that there may be plant-specific TS, which meet the 18 criteria described in the Traveler, that may be included in the RICT Program scope.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 26 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Division 1 or 2 125 VDC electrical power subsystem inoperable for reasons other than Conditions B or C.
| |
| Restore Division 1 or 2 125 VDC ------------------ 3.8.4.D.1 Yes Not included in NUREG-1433 markup in TSTF-505, electrical power subsystem to (similar to but desired for inclusion in the QCNPS RICT OPERABLE status. 3.8.4.C.1) Program. This is acceptable because the Traveler states that there may be plant-specific TS, which meet the 18 criteria described in the Traveler, that may be included in the RICT Program scope.
| |
| Opposite unit 125 VDC electrical power subsystem inoperable.
| |
| Restore the opposite unit 125 --------------- 3.8.4.E1 Yes Not included in NUREG-1433 markup in TSTF-505, VDC electrical power subsystem but desired for inclusion in the QCNPS RICT to OPERABLE status. Program. This is acceptable because the Traveler states that there may be plant-specific TS, which meet the 18 criteria described in the Traveler, that may be included in the RICT Program scope.
| |
| Inverters - Operating 3.8.7 --------------------------
| |
| One [required] inverter inoperable. 3.8.7.A -------------------------- No The QCNPS TS do not contain this TS, therefore, TSTF-505 changes are not incorporated.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 27 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| Distribution Systems - 3.8.9 3.8.7 QCNPS TS numbering is different than Operating NUREG-1433.
| |
| Although the Technical Specifications numbering between the NUREG-1433 and QCNPS is different, this is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
| |
| One or more AC electrical power distribution subsystems inoperable.
| |
| 3.8.9.A.1 3.8.7.A.1 Yes TSTF-505 changes are incorporated with numbering Restore AC electrical power difference previously noted.
| |
| distribution subsystem(s) to OPERABLE status.
| |
| However, under certain circumstances, with more than one AC distribution subsystem inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT in this circumstance.
| |
| [One or more AC vital buses 3.8.9.B.1 --------------------------- No The QCNPS TS do not contain this TS, therefore, inoperable.] TSTF-505 changes are not incorporated.
| |
| | |
| License Amendment Request Attachment 4 Adopt Risk Informed Completion Times TSTF-505 Page 28 of 28 Docket Nos. 50-254 and 50-265 Cross-Reference of TSTF-505 and QCNPS Technical Specifications Tech Spec Description TSTF-505 QCNPS Tech Spec Apply Comments Tech Spec RICT?
| |
| One or more [station service] DC electrical power distribution subsystems inoperable.
| |
| Restore DC electrical power 3.8.9.C.1 3.8.7.B.1 Yes TSTF-505 changes are incorporated with numbering distribution subsystem(s) to difference previously noted.
| |
| OPERABLE status.
| |
| One or more required opposite unit AC or DC electrical power distribution subsystems inoperable.
| |
| . 3.8.7.C.1 Yes Not included in NUREG-1433 markup in TSTF-505, Restore required opposite unit AC but desired for inclusion in the QCNPS RICT or DC electrical power distribution Program. This is acceptable because the Traveler subsystems to OPERABLE status. states that there may be plant-specific TS, which meet the 18 criteria described in the Traveler, that may be included in the RICT Program scope.
| |
| 5.5 5.5 Programs and Manuals Programs and Manuals 5.5.15 [NEW TS] 5.5.15 The QCNPS TS do not currently contain this program. The new RICT Program will be added to the QCNPS TS 5.5.15 consistent with TSTF-505.
| |
| | |
| ATTACHMENT 5 Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" RICT Program Implementation Items
| |
| | |
| License Amendment Request Attachment 5 Adopt Risk Informed Completion Times TSTF-505 Page 1 of 1 Docket Nos. 50-254 and 50-265 RICT Program Implementation Items The table below identifies the items that are required to be completed prior to implementation of the Risk Informed Completion Time (RICT) Program at Quad Cities Nuclear Power Station (QCNPS). All issues identified below will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are Probabilistic Risk Assessment (PRA) upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009), and any findings will be resolved and reflected in the PRA of record prior to the implementation of the RICT Program.
| |
| Table A5-1 QCNPS RICT Program Implementation Items Source Description Implementation Item The structures, systems, and One or more lines with one components (SSCs) are not Enclosure 1, Table E1-1, reactor building-to- modeled. The model will be Technical Specification (TS) suppression chamber updated to include these 3.6.1.7.A vacuum breaker not closed. SSCs prior to exercising the RICT program for this TS.
| |
| One line with one or more The SSCs are not modeled.
| |
| Enclosure 1, Table E1-1, reactor building-to- The model will be updated to Technical Specification (TS) suppression chamber include these SSCs prior to 3.6.1.7.C vacuum breakers inoperable exercising the RICT program for opening. for this TS.
| |
| The SSCs are not modeled.
| |
| Two lines with one reactor Enclosure 1, Table E1-1, The model will be updated to building-to-suppression Technical Specification (TS) include these SSCs prior to chamber vacuum breakers 3.6.1.7.E exercising the RICT program inoperable for opening.
| |
| for this TS.
| |
| Complete EC 636914, Enclosure 4, Section 5.2.3.3, LIP barriers are modified to Update to LIP Barriers to LIP Flood Barrier Upgrades protect the plant up to 599.0' Assist the Station External and Deployment Flood, modifications.
| |
| | |
| ATTACHMENT 6 Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Proposed Renewed Facility Operating License Changes (Mark-ups)
| |
| | |
| License Amendment Request Attachment 6 Adopt Risk Informed Completion Times TSTF-505 Page 1 of 1 Docket Nos. 50-254 and 50-265 INSERT for Unit 1 and Unit 2 Operating License Adoption of Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extension Completion Times - RITSTF Initiative 4b" Constellation is approved to implement TSTF-505, Revision 2, modifying the Technical Specifications requirements related to Completion Times (CT) for Required Actions to provide the option to calculate a longer, risk-informed CT (RICT). The methodology for using the new Risk Informed Completion Time Program is described in NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," Revision 0, which was approved by the NRC on May 17, 2007.
| |
| Constellation will complete the implementation items listed in Attachment 5 of Constellation Letter to the NRC dated June 8, 2023, prior to implementation of the RICT Program. All issues identified in Attachment 5 will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to the implementation of the RICT Program.
| |
| | |
| ENCLOSURE 1 License Amendment Request Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" List of Revised Required Actions to Corresponding PRA Functions
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions
| |
| : 1. Introduction Section 4.0, Item 2 of the U.S. Nuclear Regulatory Commission (NRC) Final Safety Evaluation (Reference [1]) for NEI 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines" (Reference [2]), identifies the following needed content:
| |
| The License Amendment Request (LAR) will provide identification of the Technical Specifications (TS) Limiting Conditions for Operation (LCOs) and action requirements to which the RMTS will apply.
| |
| The LAR will provide a comparison of the TS functions to the probabilistic risk assessment (PRA) modeled functions of the structures, systems, and components (SSCs) subject to those LCO actions.
| |
| The comparison should justify that the scope of the PRA model, including applicable success criteria such as number of SSCs required, flow rate, etc., are consistent with licensing basis assumptions (i.e., 50.46 Emergency Core Cooling System (ECCS) flowrates) for each of the TS requirements, or an appropriate disposition or programmatic restriction will be provided.
| |
| : 2. Revised Required Actions to Corresponding PRA Functions This enclosure provides confirmation that that the Quad Cities Nuclear Power Station (QCNPS)
| |
| PRA models include the necessary scope of SSCs and their functions to address each proposed application of the Risk Informed Completion Time (RICT) Program to the proposed scope TS LCO Conditions, and provides the information requested by Section 4.0, Item 2 of the NRC Final Safety Evaluation. The comparison includes each of the TS LCO Conditions and associated Required Actions within the scope of the RICT Program. The QCNPS PRA model has the capability to model directly or through use of a bounding surrogate the risk impact of entering each of the TS LCOs in the scope of the RICT Program.
| |
| Table E1-1 below lists each TS LCO Condition to which the RICT Program is proposed to be applied and documents the following information regarding the TSs with the associated safety analyses, the analogous PRA functions, and the results of the comparison:
| |
| TS Condition: Lists the alphanumerical TS and Condition.
| |
| TS Condition
| |
| | |
| == Description:==
| |
| Lists the LCOs and Condition statements within the scope of the RICT Program.
| |
| SSCs Covered by TS LCO: Lists the SSCs addressed by each action requirement.
| |
| SSCs Modeled in PRA: Indicates whether the SSCs addressed by the TS LCO Condition are included in the PRA.
| |
| Function Covered by TS LCO: Summarizes the required functions from the design basis analyses.
| |
| Design Success Criteria: Summarizes the success criteria from the design basis analyses.
| |
| PRA Success Criteria: The function success criteria modeled in the PRA.
| |
| Other Comments: Provides the justification or resolution to address any inconsistencies between the TS and PRA functions regarding the scope of SSCs and the success E1-2
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions criteria. Where the PRA scope of SSCs is not consistent with the TS, additional information is provided to describe how the LCO Condition can be evaluated using appropriate surrogate events. Differences in the success criteria for TS functions are addressed to demonstrate the PRA criteria provide a realistic estimate of the risk of the TS Condition as required by NEI 06-09, Revision 0-A.
| |
| The corresponding SSCs for each TS LCO and the associated TS functions are identified and compared to the PRA. This description also includes the design success criteria and the applicable PRA success criteria. Any differences between the scope or success criteria are described in Table E1-1. Scope differences are justified by identifying appropriate surrogate events which permit a risk evaluation to be completed using the Real-Time Risk (RTR) tool for the RICT program. Differences in success criteria typically arise due to the requirement in the PRA standard to make PRAs realistic rather than bounding, whereas design basis criteria are necessarily conservative and bounding. The use of realistic success criteria is necessary to conform to capability Category II of the PRA standard (Reference [3]) as required by NEI 06-09, Revision 0-A.
| |
| E1-3
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Provide a backup SSCs modeled in capability for One Standby PRA consistent with bringing the Liquid Control Same as Design TS scope and can be 3.1.7.A Two SLC trains Yes reactor from full One of two trains (SLC) subsystem Success Criteria explicitly included in power to a cold, inoperable. the RTR tool for the Xenon-free RICT program.
| |
| shutdown.
| |
| Individual RPS instrumentation inputs One of two to the RPS logic channels, taken system are not Reactor Protection twice modeled in the PRA.
| |
| One or more Provide reactor System (RPS) required Not trip signal based Same as Design 3.3.1.1.A Instrumentation See Section 5.1 Common cause channels explicitly on plant Success Criteria outlined in for a detailed failure of the RPS inoperable. parameters.
| |
| Table 3.3.1.1-1 RPS signal electrical system is diversity used as a discussion. conservative surrogate for failure of the RPS.
| |
| E1-4
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Individual RPS instrumentation inputs One of two to the RPS logic One or more channels, taken system are not Functions with twice modeled in the PRA.
| |
| RPS Provide reactor one or more Instrumentation Not trip signal based Same as Design 3.3.1.1.B required See Section 5.1 Common cause outlined in explicitly on plant Success Criteria channels for a detailed failure of the RPS Table 3.3.1.1-1 parameters.
| |
| inoperable in RPS signal electrical system is both trip systems. diversity used as a discussion. conservative surrogate for failure of the RPS.
| |
| Common cause One or more Two of four failure of high level Feedwater Feedwater Reactor high level channels feedwater trip failure System and main System and main Feedwater System will be used as a Not turbine trip to Same as Design 3.3.2.2.A turbine high and main turbine See Section 5.2 conservative explicitly prevent water Success Criteria water level trip trip for a detailed surrogate for failure of intrusion into channels instrumentation diversity any channel of turbines inoperable. discussion. feedwater instrumentation.
| |
| E1-5
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Anticipated Transient Without Scram Recirculation One of two Pump Trip channels Failure of recirculation (ATWS-RPT) pump breakers will be One or more System includes Not Recirculation See Section 5.3 Same as Design used as a surrogate 3.3.4.1.A channels sensors, relays, explicitly pump trip for a detailed Success Criteria for failure of the inoperable. bypass capability, ATWS-RPT ATWS-RPT circuit breakers, signal diversity instrumentation and switches that discussion.
| |
| are necessary to cause initiation of a recirculation pump trip E1-6
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Emergency Core Success of the Cooling System ECCS (ECCS) actuation instrumentation is instrumentation for dependent upon Core Spray (CS), Yes (1.a, individual SSCs modeled in As required by Low Pressure 1.b, 2.a, instrumentation PRA consistent with Required Action Initiate ECCS (CS, Coolant Injection 2.b, 2.d, channel functions Same as Design TS scope and can be 3.3.5.1.B A.1 and LPCI, associated (LPCI), High 2.j), not outlined in Success Criteria explicitly included in referenced in DG, HPCI)
| |
| Pressure Coolant explicitly Table 3.3.5.1-1. the RTR tool for the Table 3.3.5.1-1.
| |
| Injection (HPCI), (3.a, 3.b) RICT program.
| |
| Diesel Generators See Section 5.4 (DGs) for a detailed signal diversity See Notes 2 and 3 discussion.
| |
| Success of the ECCS instrumentation is dependent upon Yes (1.c, individual SSCs modeled in As required by ECCS actuation 1.e, 2.c, 2.e, instrumentation PRA consistent with Required Action instrumentation for Initiate ECCS (CS, 2.g, 2.h, 2.i, channel functions Same as Design TS scope and can be 3.3.5.1.C A.1 and CS, LPCI, DGs LPCI, associated 2.k), not outlined in Success Criteria explicitly included in referenced in DG) explicitly Table 3.3.5.1-1. the RTR tool for the Table 3.3.5.1-1. See Notes 2 and 3 3.c, 3.g) RICT program.
| |
| See Section 5.4 for a detailed signal diversity discussion.
| |
| E1-7
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Success of the ECCS instrumentation is dependent upon CCST level-low and individual suppression pool As required by instrumentation water level-high HPCI Required Action ECCS actuation Not channel functions Same as Design instrumentation not 3.3.5.1.D A.1 and instrumentation for Initiate HPCI explicitly outlined in Success Criteria explicitly modeled, referenced in HPCI Table 3.3.5.1-1. therefore the fail to Table 3.3.5.1-1.
| |
| start is used as See Section 5.4 surrogate.
| |
| for a detailed signal diversity discussion.
| |
| Success of the ECCS instrumentation is dependent upon ECCS pump ECCS actuation individual As required by discharge flow-low, instrumentation for instrumentation Required Action Initiate ECCS (CS, bypass, not explicitly CS, LPCI, HPCI, Not channel functions Same as Design 3.3.5.1.E A.1 and LPCI, associated modeled. Minimum DGs explicitly outlined in Success Criteria referenced in DG) flow valves fail to Table 3.3.5.1-1.
| |
| Table 3.3.5.1-1. open used as See Notes 2 and 3 surrogates.
| |
| See Section 5.4 for a detailed signal diversity discussion.
| |
| E1-8
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Success of the ECCS instrumentation is dependent upon individual ADS actuation As required by ECCS actuation instrumentation instrumentation not Required Action instrumentation for Not channel functions Same as Design explicitly modeled; 3.3.5.1.F A.1 and Automatic Initiate ADS explicitly outlined in Success Criteria relief valves fail to referenced in Depressurization Table 3.3.5.1-1. open used as Table 3.3.5.1-1. System (ADS) surrogates.
| |
| See Section 5.4 for a detailed signal diversity discussion.
| |
| Success of the ECCS instrumentation is dependent upon individual ADS actuation As required by instrumentation instrumentation not Required Action ECCS actuation Not channel functions Same as Design explicitly modeled; 3.3.5.1.G A.1 and instrumentation for Initiate ADS explicitly outlined in Success Criteria relief valves fail to referenced in ADS Table 3.3.5.1-1. open used as Table 3.3.5.1-1.
| |
| surrogates.
| |
| See Section 5.4 for a detailed signal diversity discussion.
| |
| E1-9
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Success of the RCIC instrumentation is dependent upon Reactor Vessel individual level SSCs modeled in As required by instrumentation Instrumentation PRA consistent with Required Action channel functions supporting Same as Design TS scope and can be 3.3.5.3.B A.1 and Yes Initiate RCIC outlined in Reactor Core Success Criteria explicitly included in referenced in Table 3.3.5.3-1.
| |
| Isolation Cooling the RTR tool for the Table 3.3.5.3-1.
| |
| (RCIC) automatic RICT program.
| |
| See Section 5.5 initiation for a detailed RCIC signal diversity discussion.
| |
| E1-10
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Success of the RCIC instrumentation is dependent upon individual CCST level-low and CCST and As required by instrumentation torus water level-high suppression pool Required Action channel functions RCIC instrumentation level Not Same as Design 3.3.5.3.D A.1 and Initiate RCIC outlined in not explicitly modeled, instrumentation explicitly Success Criteria referenced in Table 3.3.5.3-1. therefore fail to open supporting RCIC Table 3.3.5.3-1. for injection valves are automatic initiation See Section 5.5 used as surrogates.
| |
| for a detailed RCIC signal diversity discussion.
| |
| E1-11
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Success of the primary containment instrumentation is The logic for primary dependent upon containment isolation Primary individual is not modeled in One or more Containment Initiate closure of instrumentation detail. Therefore, a required Isolation Not primary Same as Design 3.3.6.1.A channel functions surrogate is chosen channels Instrumentation explicitly containment Success Criteria outlined in that represents a inoperable outlined in isolation valves Table 3.3.6.1-1. failure of the Table 3.3.6.1-1 containment isolation See Section 5.6 signal.
| |
| for a detailed signal diversity discussion.
| |
| Success of the Relief Valve Set Relief valve instrumentation is instrumentation is not One relief valve Relief Valve Set Mitigate dependent upon explicitly modeled, inoperable due to Instrumentation Not Same as Design 3.3.6.3.A overpressurization individual therefore fail to open inoperable outlined in explicitly Success Criteria transients instrumentation for the relief valves channel(s). Table 3.3.6.3-1 channel functions are used as outlined in surrogates.
| |
| Table 3.3.6.3-1 E1-12
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Loss of Power (LOP) Failure of DG instrumentation autostart, load Two of two trains Failure of DG includes sensors, shed, and autostart, load shed, relays, bypass undervoltage One or more Undervoltage See Section 5.7 and undervoltage capability, circuit Not relays will be 3.3.8.1.A channels functions for each for a detailed relays will be used as breakers, and explicitly used as a inoperable. 4160 V ESS bus LOP signal a surrogate for failure switches that are surrogate for diversity of the LOP necessary to trip failure of the discussion. instrumentation.
| |
| offsite power LOP circuits and start instrumentation.
| |
| the DGs Relieve pressure Five relief valves SSCs modeled in during a limiting (Four ERV and PRA consistent with event and One valve (non-One relief valve one Target Rock Four of five relief TS scope and can be 3.4.3.A Yes maintain ASME ATWS), 12 of 13 inoperable. Valve) valves explicitly included in Code limit on valves (ATWS) the RTR tool for the reactor See Note 1 RICT program.
| |
| overpressure One LPCI subsystem SSCs modeled in inoperable for Low pressure One LPCI PRA consistent with Two LPCI reasons other injection into the Two LPCI pumps, subsystem or TS scope and can be 3.5.1.B subsystems, two Yes than Condition A reactor pressure one CS pump one CS explicitly included in CS subsystems OR one Core vessel (RPV) subsystem the RTR tool for the Spray subsystem RICT program.
| |
| inoperable.
| |
| E1-13
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO SSCs modeled in One LPCI pump PRA consistent with Low pressure One LPCI pump in each Two LPCI One of two LPCI TS scope and can be 3.5.1.C Yes injection into the in each subsystem subsystems subsystems explicitly included in RPV subsystem inoperable. the RTR tool for the RICT program.
| |
| Low pressure SSCs modeled in Two LPCI injection function PRA consistent with subsystems Low pressure One train of one Two LPCI is achieved TS scope and can be 3.5.1.E inoperable for Yes injection into the low pressure subsystems through two explicitly included in reasons other RPV source redundant CS the RTR tool for the than Condition C.
| |
| subsystems RICT program.
| |
| Injection function is achieved by SSCs modeled in RCIC being PRA consistent with High pressure operable and low HPCI System Same as Design TS scope and can be 3.5.1.G HPCI System Yes injection into the pressure ECCS inoperable. Success Criteria explicitly included in RPV injection/spray the RTR tool for the subsystems in RICT program.
| |
| conjunction with ADS Provide Five ADS valves SSCs modeled in depressurization (Four ERV, one PRA consistent with of RCS during One ADS valve Target Rock Four of five Two of five TS scope and can be 3.5.1.H Yes small break LOCA inoperable. valve) valves valves explicitly included in if HPCI cannot the RTR tool for the maintain RPV See Note 1 RICT program.
| |
| level E1-14
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Provide makeup water and SSCs modeled in maintain RPV PRA consistent with level above top of RCIC System Same as Design TS scope and can be 3.5.3.A RCIC System Yes core following HPCI is operable inoperable. Success Criteria explicitly included in RPV isolation with the RTR tool for the loss of coolant RICT program.
| |
| flow from feedwater system The primary Primary One of two containment air locks containment air primary are not modeled Primary Primary lock inoperable Not containment air Maintain primary therefore a pre-3.6.1.2.C containment air containment for reasons other explicitly lock doors containment existing containment lock equipment isolation than Condition A maintain failure will be used as or B. boundary a conservative surrogate.
| |
| One or more penetration flow Minimize the loss Not all PCIVs are paths with one of reactor coolant modeled therefore a primary inventory and One of two pre-existing containment Not establish the Same as Design 3.6.1.3.A PCIVs isolation valves containment failure isolation valve explicitly primary Success Criteria per penetration will be used as a (PCIV) containment conservative inoperable for boundary during surrogate.
| |
| reasons other accidents than Condition D.
| |
| E1-15
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO The low set mode of the relief valves is not explicitly modeled in Prevents the PRA, however the excessive short failure of the relief One low set relief Two low set relief Not duration S/RV Same as Design 3.6.1.6.A One of two valves valves to open can be valve inoperable. valves explicitly cycles with valve Success Criteria used as a surrogate actuation at the since failure of the relief setpoint.
| |
| relief valves to open bounds the risk of the low set mode.
| |
| The SSCs are not modeled. The model One or more Relieve vacuum will be updated to lines with one Reactor building- when primary One of vacuum include these SSCs reactor building-to-suppression containment breakers closed prior to exercising the 3.6.1.7.A to-suppression No N/A chamber vacuum depressurizes in each of the two RICT program for this chamber vacuum breakers below reactor lines TS. This is an breaker not building pressure Implementation Item closed.
| |
| identified in Attachment 5.
| |
| E1-16
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO One line with one or more reactor building-to-Two vacuum suppression 3.6.1.7.C See TS 3.6.1.7.A breakers open in See TS 3.6.1.7.A chamber vacuum one of two paths breakers inoperable for opening.
| |
| Two lines with one or more reactor building-Two vacuum to-suppression 3.6.1.7.E See TS 3.6.1.7.A. breakers open in See TS 3.6.1.7.A.
| |
| chamber vacuum one of two paths breakers inoperable for opening.
| |
| The opening function One required of the suppression suppression chamber to drywell Suppression chamber-to- vacuum breakers is chamber-to- Not Relieve vacuum in 7 of 12 vacuum Same as Design 3.6.1.8.A drywell vacuum not modeled in the drywell vacuum explicitly the drywell breakers Success Criteria breaker PRA. The vapor breakers inoperable for suppression function opening. is modeled and is used as a surrogate.
| |
| E1-17
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO SSCs modeled in One RHR Maintain primary Two Residual PRA consistent with suppression pool containment peak Heat Removal One of four RHR One of four RHR TS scope and can be 3.6.2.3.A cooling Yes pressure and (RHR) loops loops explicitly included in subsystem temperature below subsystems the RTR tool for the inoperable. design limits RICT program.
| |
| Adequately scrub SSCs modeled in Two RHR inorganic iodines PRA consistent with One RHR drywell subsystems and particulates One of four RHR One of four RHR TS scope and can be 3.6.2.6.A spray subsystem Yes from primary loops loops explicitly included in inoperable.
| |
| See Note 4 containment the RTR tool for the atmosphere RICT program.
| |
| Provide cooling SSCs modeled in Two Residual One RHRSW water for the RHR PRA consistent with Heat Removal One of two One of two pump in each system heat TS scope and can be 3.7.1.B Service Water Yes pumps in each pumps in each subsystem exchangers explicitly included in (RHRSW) subsystem subsystem inoperable. following a DBA or the RTR tool for the subsystems transient RICT program.
| |
| One RHRSW subsystem One of two One of two 3.7.1.C inoperable for See TS 3.7.1.B. See TS 3.7.1.B.
| |
| subsystems subsystems reasons other than Condition A.
| |
| E1-18
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO SSCs modeled in Provide RPV Safe Shutdown PRA consistent with makeup during Makeup Pump Same as Design TS scope and can be 3.7.9.A SSMP Yes RPV isolation with One of one train (SSMP) System Success Criteria explicitly included in loss of feedwater inoperable the RTR tool for the flow RICT program.
| |
| Reserve Auxiliary Transformers (RATs), Unit SSCs modeled in One offsite power Auxiliary PRA consistent with One required source As needed to Transformers Supply AC loads TS scope and can be 3.8.1.A offsite circuit Yes supply supported (UATs), during operation explicitly included in inoperable. See Section 7 for functions associated the RTR tool for the additional details.
| |
| breakers, and RICT program.
| |
| offsite power supplies SSCs modeled in Two out of three PRA consistent with Supply AC loads DGs One required DG Three DGs and One out of three TS scope and can be 3.8.1.B Yes during abnormal inoperable. support systems DGs explicitly included in operation See Section 7 for the RTR tool for the additional details.
| |
| RICT program.
| |
| SSCs modeled in RATs, UATs, Two required PRA consistent with Two required associated offsite sources As needed to Supply AC loads TS scope and can be 3.8.1.C offsite circuits breakers, and Yes supply supported during operation explicitly included in inoperable. offsite power See Section 7 for functions the RTR tool for the supplies additional details.
| |
| RICT program.
| |
| E1-19
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO RATs, UATs, SSCs modeled in One required Two required associated PRA consistent with offsite circuit Supply AC loads sources breakers, offsite Same as Design TS scope and can be 3.8.1.D inoperable OR Yes during abnormal power supplies, Success Criteria explicitly included in one required DG operation See Section 7 for three DGs and the RTR tool for the inoperable. additional details.
| |
| support systems RICT program.
| |
| SSCs modeled in One out of two One 250 VDC PRA consistent with Supply DC loads subsystems electrical power Two 250 VDC Same as Design TS scope and can be 3.8.4.A Yes during normal subsystem subsystems Success Criteria explicitly included in operation See Section 7 for inoperable. the RTR tool for the additional details.
| |
| RICT program.
| |
| Division 1 or 2 SSCs modeled in One out of two 125 VDC battery PRA consistent with Division 1 and Supply DC loads subsystems inoperable as a Same as Design TS scope and can be 3.8.4.B Division 2 125 Yes during normal result of Success Criteria explicitly included in VDC batteries operation See Section 7 for maintenance or the RTR tool for the additional details.
| |
| testing. RICT program.
| |
| Division 1 or 2 125 VDC battery SSCs modeled in inoperable, due One out of two PRA consistent with to the need to Division 1 and Supply DC loads subsystems Same as Design TS scope and can be 3.8.4.C replace the Division 2 125 Yes during normal Success Criteria explicitly included in battery, as VDC batteries operation See Section 7 for the RTR tool for the determined by additional details.
| |
| RICT program.
| |
| maintenance or testing.
| |
| E1-20
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Division 1 or 2 125 VDC SSCs modeled in One out of two electrical power PRA consistent with Division 1 and Supply DC loads subsystems subsystem Same as Design TS scope and can be 3.8.4.D Division 2 125 Yes during normal inoperable for Success Criteria explicitly included in VDC subsystem operation See Section 7 for reasons other the RTR tool for the additional details.
| |
| than Condition B RICT program.
| |
| or C.
| |
| SSCs modeled in One out of one Opposite unit 125 PRA consistent with Opposite unit Supply DC loads subsystem VDC electrical Same as Design TS scope and can be 3.8.4.E Division 2 Yes during normal power subsystem Success Criteria explicitly included in subsystem operation See Section 7 for inoperable. the RTR tool for the additional details.
| |
| RICT program.
| |
| Provide AC power Two AC SSCs modeled in One or more AC distribution to One of two distribution PRA consistent with electrical power required divisional subsystems subsystems, Same as Design TS scope and can be 3.8.7.A distribution Yes loads to shut associated buses, Success Criteria explicitly included in subsystems down reactor and See Section 7 for MCCS, and the RTR tool for the inoperable. maintain in safe additional details.
| |
| distribution panels RICT program.
| |
| condition Provide DC power One of two SSCs modeled in One or more DC Two 125 VDC distribution to subsystems for PRA consistent with electrical power distribution required divisional 125 and 250 VDC Same as Design TS scope and can be 3.8.7.B distribution subsystems, two Yes loads to shut Success Criteria explicitly included in subsystems 250 VDC down reactor and See Section 7 for the RTR tool for the inoperable. subsystems maintain in safe additional details. RICT program.
| |
| condition E1-21
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Function TS TS Condition SSCs Covered by Design Success PRA Success Modeled in Covered by TS Other Comments Condition Description TS LCO Criteria Criteria PRA? LCO Two AC distribution One of two One or more Provide AC/DC subsystems, subsystems for SSCs modeled in required opposite power distribution associated buses, each AC/DC PRA consistent with unit AC or DC to required MCCS, distribution distribution Same as Design TS scope and can be 3.8.7.C electrical power Yes divisional loads to panels, two 125 system Success Criteria explicitly included in distribution shut down reactor VDC distribution the RTR tool for the subsystems and maintain in subsystems, two See Section 7 for RICT program.
| |
| inoperable. safe condition 250 VDC additional details.
| |
| subsystems Table E1-1 Notes:
| |
| : 1. The five relief valves consist of four electromagnetic relief valves (ERV) and the Target Rock safety/relief valve. The Target Rock valve performs both relief valve and safety valve functions and is credited as one of each valve.
| |
| : 2. QCNPS has two independent core spray divisions (subsystems or loop). Each division consists of a 4500 gal/min capacity pump, valves, piping and an independent circular sparger ring inside the inner shroud just over the core. Suction water is supplied by the suppression pool.
| |
| : 3. LPCI is a functional mode of the RHR system. LPCI has two divisions (subsystems or loop) which consists of a heat exchanger, two RHR pumps in parallel, and associated piping. The two divisions of LPCI are cross connected by a single header, making it possible to supply either division from the pumps in the other division.
| |
| : 4. Each of the two RHR drywell spray subsystems contains two pumps, one heat exchanger, drywell spray valves, and a spray header inside the drywell. Each RHR drywell spray subsystem is capable of recirculating water from the RHR suppression pool through a heat exchanger and dispersed through the RHR drywell spray nozzles. The LOCA radiological dose analysis credits the RHR drywell spray system for scrubbing radionuclides from the drywell air space.
| |
| E1-22
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions
| |
| : 3. Example RICT Calculations Examples of calculated RICTs are provided in Table E1-2 for each individual condition to which the RICT applies (assuming no other SSCs modeled in the PRA are unavailable). Following TSTF-505 implementation, the actual RICT values will be calculated using the actual plant configuration and the current revision of the PRA models representing the as-built, as-operated condition of the plant, as required by NEI 06-09, Revision 0-A and the NRC safety evaluation, and may differ from the RICTs presented in Table E1-2.
| |
| Table E1-2 below lists the calculated RICTs for each TS condition using the method outlined in NEI 06-09, Revision 0-A (Reference [2]), shown below. The same equation was used to calculate the Large Early Release Frequency (LERF) RICT by simply using the RICT Incremental Conditional Large Early Release Probability (ICLERP) Limit and LERF instead.
| |
| 365 1
| |
| The RICT Incremental Conditional Core Damage Probability (ICCDP) limit is 1.00E-05, while the RICT ICLERP limit is 1.00E-06. The RICTs are limited to a maximum of thirty (30) days, and to a minimum of the original TS completion time.
| |
| Table E1-2 provides example RICT calculations for the purposes of this Enclosure.
| |
| Table E1-2: In-Scope TS/LCO Conditions RICT Estimate RICT TS LCO Condition Estimate1,2 Standby Liquid Control (SLC) System - One SLC subsystem 3.1.7.A 30.0 inoperable.
| |
| Reactor Protection System (RPS) Instrumentation - One or 3.3.1.1.A 2.3 more required channels inoperable.
| |
| Reactor Protection System (RPS) Instrumentation - One or 3.3.1.1.B more Functions with one or more required channels 2.3 inoperable in both trip systems.
| |
| Feedwater System and Main Turbine High Water Level Trip 3.3.2.2.A Instrumentation - One or more Feedwater System and main 30.0 turbine high water level trip channels inoperable.
| |
| Anticipated Transient Without Scran Recirculation Pump Trip 3.3.4.1.A (ATWS-RPT) Instrumentation - One or more channels 30.0 inoperable.
| |
| Emergency Core Cooling System (ECCS) Instrumentation -
| |
| 3.3.5.1.B As required by Required Action A.1 and referenced in Table 30.0 3.3.5.1-1.
| |
| ECCS Instrumentation - As required by Required Action A.1 3.3.5.1.C 30.0 and referenced in Table 3.3.5.1-1.
| |
| E1-23
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-2: In-Scope TS/LCO Conditions RICT Estimate RICT TS LCO Condition Estimate1,2 ECCS Instrumentation - As required by Required Action A.1 3.3.5.1.D 30.0 and referenced in Table 3.3.5.1-1.
| |
| ECCS Instrumentation - As required by Required Action A.1 3.3.5.1.E 30.0 and referenced in Table 3.3.5.1-1.
| |
| ECCS Instrumentation - As required by Required Action A.1 3.3.5.1.F 30.0 and referenced in Table 3.3.5.1-1.
| |
| ECCS Instrumentation - As required by Required Action A.1 3.3.5.1.G 30.0 and referenced in Table 3.3.5.1-1.
| |
| Reactor Core Isolation Cooling (RCIC) System 3.3.5.3.B Instrumentation - As required by Required Action A.1 and 24.6 referenced in Table 3.3.5.3-1.
| |
| RCIC System Instrumentation - As required by Required 3.3.5.3.D 30.0 Action A.1 and referenced in Table 3.3.5.3-1.
| |
| Primary Containment Isolation Instrumentation - One or 3.3.6.1.A 30.0 more required channels inoperable Relief Valve Instrumentation - One relief valve inoperable 3.3.6.3.A 30.0 due to inoperable channel(s)
| |
| Loss of Power (LOP) Instrumentation - One or more 3.3.8.1.A 30.0 channels inoperable.
| |
| 3.4.3.A Safety and Relief Valves - One relief valve inoperable. 30.0 ECCS-Operating - One LPCI subsystem inoperable for 3.5.1.B reasons other than Condition A OR one Core Spray 16.1 subsystem inoperable.
| |
| ECCS-Operating - One LPCI pump in each subsystem 3.5.1.C 30.0 inoperable.
| |
| ECCS-Operating - Two LPCI subsystems inoperable for 3.5.1.E 30.0 reasons other than Condition C.
| |
| 3.5.1.G ECCS-Operating - HPCI System inoperable. 30.0 3.5.1.H ECCS-Operating - One ADS valve inoperable. 30.0 3.5.3.A RCIC System - RCIC System inoperable. 30.0 Primary Containment Air Lock - Primary containment air 3.6.1.2.C 6.0 lock inoperable for reasons other than Condition A or B.
| |
| Primary Containment Isolation Valves (PCIVs) - One or 3.6.1.3.A more penetration flow paths with one PCIV inoperable for 6.0 reasons other than Condition D.
| |
| 3.6.1.6.A Low set relief valves - One low set relief valve inoperable. 30.0 Reactor Building-to-Suppression Chamber Vacuum 3.6.1.7.A Breakers - One or more lines with one reactor building-to- N/A suppression chamber vacuum breaker not closed.
| |
| E1-24
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-2: In-Scope TS/LCO Conditions RICT Estimate RICT TS LCO Condition Estimate1,2 Reactor Building-to-Suppression Chamber Vacuum Breakers - One line with one or more reactor building-to-3.6.1.7.C N/A suppression chamber vacuum breakers inoperable for opening.
| |
| Reactor Building-to-Suppression Chamber Vacuum Breakers - Two lines with one or more reactor building-to-3.6.1.7.E N/A suppression chamber vacuum breakers inoperable for opening.
| |
| Suppression Chamber-to-Drywell Vacuum Breakers - One 3.6.1.8.A required suppression chamber-to-drywell vacuum breaker 3.5 inoperable for opening.
| |
| Residual Heat Removal (RHR) Suppression Pool Cooling -
| |
| 3.6.2.3.A 30.0 One RHR suppression pool cooling subsystem inoperable.
| |
| RHR Drywell Spray - One RHR drywell spray subsystem 3.6.2.6.A 30.0 inoperable.
| |
| Residual Heat Removal Service Water (RHRSW) System -
| |
| 3.7.1.B 30.0 One RHRSW pump in each subsystem inoperable.
| |
| RHRSW System - One RHRSW subsystem inoperable for 3.7.1.C 24.8 reasons other than Condition A.
| |
| Safe Shutdown Makeup Pump (SSMP) System - SSMP 3.7.9.A 30.0 System inoperable AC Sources-Operating - One required offsite circuit 3.8.1.A 14.5 inoperable.
| |
| 3.8.1.B AC Sources-Operating - One required DG inoperable. 30.0 AC Sources-Operating - Two required offsite circuits 3.8.1.C 14.4 inoperable.
| |
| AC Sources-Operating - One required offsite circuit 3.8.1.D 4.8 inoperable AND one required DG inoperable.
| |
| DC Sources-Operating - One 250 VDC electrical power 3.8.4.A 30.0 subsystem inoperable.
| |
| DC Sources-Operating - Division 1 or 2 125 VDC battery 3.8.4.B 14.8 inoperable as a result of maintenance or testing.
| |
| DC Sources-Operating - Division 1 or 2 125 VDC battery 3.8.4.C inoperable, due to the need to replace the battery, as 14.8 determined by maintenance or testing.
| |
| DC Sources-Operating - Division 1 or 2 125 VDC electrical 3.8.4.D power subsystem inoperable for reasons other than 5.4 Condition B or C.
| |
| DC Sources-Operating - Opposite unit 125 VDC electrical 3.8.4.E 5.5 power subsystem inoperable.
| |
| Distribution Systems-Operating - One or more AC electrical 3.8.7.A 1.3 power distribution subsystems inoperable.
| |
| E1-25
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-2: In-Scope TS/LCO Conditions RICT Estimate RICT TS LCO Condition Estimate1,2 Distribution Systems-Operating - One or more DC electrical 3.8.7.B 5.4 power distribution subsystems inoperable.
| |
| Distribution Systems-Operating - One or more required 3.8.7.C opposite unit AC or DC electrical power distribution 5.5 subsystems inoperable.
| |
| Table E1-2 Notes:
| |
| : 1. RICTs are based on the Unit 1 internal events and internal fire PRA model calculations with seismic and high winds core damage frequency (CDF) and large early release frequency (LERF) penalties. RICTs calculated to be greater than 30 days are capped at 30 days based on NEI 06-09-A. RICTs are rounded to nearest tenth of a day.
| |
| : 2. Per NEI 06-09, for cases where the total CDF or LERF is greater than 1E-03/yr or 1E-04/yr, respectively, the RICT Program will not be entered.
| |
| : 4. Additional Justification for Specific Actions Table E1-3 lists the TSTF-505, Revision 2, Table 1 TS that require additional justification along with a description of how the additional justification is provided.
| |
| Table E1-3: TSTF-505, Revision 2, Table 1 TS that Require Additional Justification TSTF-505 QCNPS TS Description Additional Justification TS TS Source Range Monitor (SRM)
| |
| Instrumentation - One or more required SRMs inoperable in N/A - TSTF-505 changes are 3.3.1.2.A 3.3.1.2.A MODE 2 with intermediate excluded.
| |
| range monitors (IRMs) on Range 2 or below.
| |
| The QCNPS TS for 3.3.2.2.A are for, One or more feedwater and main turbine Feedwater and Main Turbine high water level trip channels High Water Level Trip inoperable. Under certain Instrumentation - Two or more circumstances, with more than 3.3.2.2.B 3.3.2.2.A feedwater and main turbine one channel inoperable, a loss high water level trip channels of function may occur.
| |
| inoperable. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
| |
| E1-26
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-3: TSTF-505, Revision 2, Table 1 TS that Require Additional Justification TSTF-505 QCNPS TS Description Additional Justification TS TS End of Cycle Recirculation N/A - The QCNPS TS do not Pump Trip (EOC-RPT) contain this TS. Therefore, a 3.3.4.1.A --
| |
| Instrumentation - One or more change is not proposed to the required channels inoperable. QCNPS TS.
| |
| The QCNPS TS for 3.3.6.3.A is for "One relief valve inoperable due to inoperable channel(s).
| |
| TS 3.3.6.3.A is bounded by Low-Low-Set (LLS) TS 3.4.3.A; therefore, the relief 3.3.6.3.A 3.3.6.3.A Instrumentation valve fail to open representation in the PRAs are used as conservative surrogates for representing the TS function.
| |
| TSTF-505 changes are incorporated. However, under certain circumstances, with more than one channel Loss of Power (LOP) inoperable, a loss of function Instrumentation - One or more 3.3.8.1.A 3.3.8.1.A may occur. Therefore, a Note channels inoperable.
| |
| is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
| |
| TSTF-505 changes are incorporated. However, under certain circumstances, with more than one primary Primary Containment Air Lock - containment airlock inoperable, Primary containment air lock excessive leakage or a loss of 3.6.1.2.C 3.6.1.2.C inoperable for reasons other function may occur. Therefore, than Condition A or B. a Note is added to the Completion Time which prohibits applying a RICT when leakage exceeds limits or there is a loss of function.
| |
| Primary Containment Isolation Valves (PCIVs) - One or more N/A - The QCNPS TS do not penetration flow paths with one contain this TS. Therefore, a 3.6.1.3.E 3.6.1.3.E or more containment purge change is not proposed to the valves not within purge valve QCNPS TS.
| |
| leakage limits.
| |
| E1-27
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-3: TSTF-505, Revision 2, Table 1 TS that Require Additional Justification TSTF-505 QCNPS TS Description Additional Justification TS TS The QCNPS TS for 3.6.1.7.E are for, Two lines with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening. This is an implementation item identified Reactor Building-to- in Attachment 5. The model Suppression will be updated to include these Chamber Vacuum Breakers - SSCs prior to exercising the Two or more lines with one or RICT program for this TS.
| |
| 3.6.1.7.D 3.6.1.7.E more reactor building-to-suppression chamber vacuum Under certain configurations breakers inoperable for with one or more reactor opening. building-to-suppression chamber vacuum breakers on two or more lines, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when the function is not maintained.
| |
| Main Turbine Bypass System -
| |
| Requirements of the LCO not N/A - TSTF-505 changes are 3.7.7.A 3.7.7.A met or Main Turbine Bypass excluded.
| |
| System inoperable.
| |
| : 5. Evaluation of Instrumentation and Control Systems The following Instrumentation TS Sections are included in the TSTF-505 application for QCNPS:
| |
| TS 3.3.1.1 - Reactor Protection System (RPS) Instrumentation TS 3.3.2.2 - Feedwater System and Main Turbine High Water Level Trip Instrumentation TS 3.3.4.1 - Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT)
| |
| Instrumentation TS 3.3.5.1 - Emergency Core Cooling System (ECCS) Instrumentation TS 3.3.5.3 - Reactor Core Isolation Cooling (RCIC) System Instrumentation TS 3.3.6.1 - Primary Containment Isolation Instrumentation TS 3.3.8.1 - Loss of Power (LOP) Instrumentation QCNPS TS Section 3.3 LCOs were developed to assure that QCNPS maintains necessary redundancy and diversity, compliant with the intent of "single failure" design criterion as defined E1-28
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions in IEEE-279-1968 and diversity requirements as defined in Topical Report NEDO-10139, "Compliance of Protection Systems to Industry Criteria: GE BWR Nuclear Steam Supply System."
| |
| 5.1 TS 3.3.1.1 - RPS Instrumentation The RPS Instrumentation employs diversity in the number and variety of different inputs which will actuate the associated equipment. The RPS, as described in the QCNPS Updated Final Safety Analysis Report (UFSAR), Section 7.2s, includes sensors, relays, bypass circuits, and switches that are necessary to cause initiation of a reactor scram. Functional diversity is provided by monitoring a wide range of dependent and independent parameters. The input parameters to the scram logic are from instrumentation that monitors reactor vessel water level, reactor vessel pressure, neutron flux, main steam line isolation valve position, turbine control valve fast closure oil pressure, turbine stop valve position, drywell pressure, and scram discharge volume water level, as well as reactor mode switch in shutdown position, manual scram signals, and RPS channel test switch scram signals. There are at least four independent sensor input signals from each of these parameters (except for the reactor mode switch in shutdown and manual scram signals). Some channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an RPS trip signal to the trip logic. Table E1-4 below presents the TS 3.3.1.1 logic descriptions for all the functions listed in TS Table 3.3.1.1-1.
| |
| Table E1-4: RPS Instrumentation Diversity Function Logic Logic Description Intermediate Range Monitors (IRMs)
| |
| The IRM System is divided into two trip systems, with four IRM channels inputting to each trip system.
| |
| One channel in each trip system is allowed to be 1.a - Neutron Flux-High 2/8 bypassed. One IRM channel tripped in each RPS trip system causes a SCRAM. The eight channels are arranged in a one-out-of-four taken twice logic 1.b - Inop 2/8 See Function 1.a.
| |
| Average Power Range Monitors (APRMs)
| |
| There are six channels of APRM. Three of the APRM channels provide trip inputs to one RPS trip system, and the other three APRM channels feed the other RPS trip system. The system is designed to allow one channel in each trip system to be 2.a - Neutron FluxHigh, 2/6 bypassed. Any one APRM channel in a trip system Setdown can cause the associated trip system to trip. Both trip systems must trip for a reactor scram in what is effectively a one-out-of-two taken twice logic (or one-out-of-three taken twice if no APRMs are bypassed).
| |
| E1-29
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-4: RPS Instrumentation Diversity Function Logic Logic Description 2.b - Flow Biased 2/6 See Function 2.a above.
| |
| Neutron FluxHigh 2.c - Fixed Neutron Flux-2/6 See Function 2.a above.
| |
| High See Function 2.a above.
| |
| For any APRM, anytime its APRM mode switch is moved to any position other than "Operate," an APRM module is unplugged, or the APRM has too few LPRM inputs (< 50%), an inoperative trip signal 2.d - Inop 2/6 will be received by the RPS, unless the APRM is bypassed. Since only one APRM in each trip system may be bypassed, only one APRM in each trip system may be inoperable without resulting in an RPS trip signal.
| |
| Four channels, with two channels in each RPS trip 3 - Reactor Vessel Steam 2/4 system, arranged in a one-out-of-two taken twice Dome PressureHigh logic.
| |
| Four channels, with two channels in each RPS trip 4 - Reactor Vessel Water 2/4 system, are arranged in a one-out-of-two taken LevelLow twice logic.
| |
| Each of the eight MSIVs has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Thus, each RPS trip 5 - Main Steam Isolation system receives an input from eight Main Steam 6 / 16 ValveClosure Isolation Valve - Closure channels. The logic is arranged such that either the inboard or outboard valve on three or more of the main steam lines must be < 90% open in order for a SCRAM to occur.
| |
| Four channels, with two channels in each RPS trip 6 - Drywell Pressure 2/4 system, arranged in a one-out-of-two taken twice High logic.
| |
| Scram Discharge Volume Water LevelHigh Scram discharge volume (SDV) high water level inputs to the RPS are from two float-type and two differential pressure-type level sensors on each of the SDVs. They are arranged such that a float-type and a differential pressure-type level sensor for 7.a - Float Switch 2/4 each channel are connected to each SDV. An actuation of any level switch causes a channel trip; an actuation of two level switches, one in each trip system, causes a scram (one-out-of-two taken twice logic).
| |
| 7.b - Differential Pressure 2/4 See Function 7.a above.
| |
| Switch E1-30
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-4: RPS Instrumentation Diversity Function Logic Logic Description Signals are initiated from position switches located on each of the four Turbine Stop Valves (TSV). A position switch with two independent contacts is associated with each TSV; one of the two switch contacts provides input to RPS trip system A; the other to RPS trip system B. Thus, each RPS trip 8 - Turbine Stop Valve-6/8 system receives an input from four Turbine Stop Closure Valve - Closure channels, each consisting of one position switch (which is common to a channel in the other RPS trip system) and a switch contact. The logic for the Turbine Stop Valve - Closure function is such that three or more TSVs must be < 90% open to produce a scram.
| |
| Four channels of Turbine Control Valve Fast 9 - Turbine Control Valve Closure, Trip Oil Pressure - Low function, with two Fast Closure, Trip Oil 2/4 channels in each trip system arranged in a one-out-PressureLow of-two taken twice logic.
| |
| Four channels of Turbine Condenser Vacuum - Low 10 - Turbine Condenser 2/4 function, with two channels in each trip system Vacuum-Low arranged in a one-out-of-two logic taken twice logic.
| |
| Each RPS trip system contains one Manual Scram logic channel that is redundant to the two automatic trip channels. Actuation of both Manual Scram logic 11 - Reactor Mode channels will result in a full reactor scram. The Switch Shutdown 2/2 Reactor Mode Switch is a single mechanical switch Position which provides a direct input into both Manual Scram channels when placed in the Shutdown position.
| |
| Each RPS Trip System contains one Manual Scram logic channel that is redundant to the two automatic trip channels. Actuation of both Manual Scram logic 12 - Manual Scram 2/2 channels will result in a full reactor scram. There is one Manual Scram push button channel for each of the RPS manual scram logic channels 5.2 TS 3.3.2.2 - Feedwater Pump and Main Turbine High Water Level Trip Instrumentation The Feedwater Pump and Main Turbine High Water Level Trip Instrumentation also employs diversity in the number and variety of different inputs which will actuate the associated equipment. The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a feedwater pump and main turbine trip signal to the trip logic. Table E1-5 below presents the logic descriptions for the functions in TS 3.3.2.2.
| |
| E1-31
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-5: Feedwater Pump and Main Turbine High Water Level Trip Instrumentation Diversity Function Logic Logic Description There are two independent high water level trip systems each containing two channels of high water level. The outputs of the channels in a trip system Reactor Vessel Water 2/4 are combined in a one-out-of-two taken twice logic LevelHigh so that both trip systems must trip to result in an initiation logic that trips the three feedwater pumps and the main turbine.
| |
| 5.3 TS 3.3.4.1 - ATWS-RPT Instrumentation The ATWS-RPT Instrumentation also employs diversity in the number and variety of different inputs which will actuate the associated equipment. The ATWS-RPT System, as described in UFSAR Section 7.8, includes sensors, relays, bypass capability, circuit breakers, and switches that are necessary to cause initiation of an RPT. The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal to the trip logic. Table E1-6 below presents the logic descriptions for the functions in TS 3.3.4.1.
| |
| Table E1-6: ATWS-RPT Instrumentation Diversity Function Logic Logic Description The ATWS-RPT trip logic consists of two independent trip systems, with two channels of Reactor Steam Dome Pressure - High and two channels of Reactor Vessel Water Level - Low Low Reactor Vessel Water in each trip system. Each ATWS-RPT trip system is 2/4 LevelLow Low a two-out-of-two logic for each function. Thus, either two Reactor Vessel Water Level - Low Low or two Reactor Pressure - High signals are needed to trip a trip system. The output of either trip system will trip both recirculation pumps motor breakers.
| |
| The ATWS-RPT trip logic consists of two independent trip systems, with two channels of Reactor Steam Dome Pressure - High and two channels of Reactor Vessel Water Level - Low Low Reactor Vessel Steam in each trip system. Each ATWS-RPT trip system is 2/4 Dome PressureHigh a two-out-of-two logic for each function. Thus, either two Reactor Vessel Water Level - Low Low or two Reactor Pressure - High signals are needed to trip a trip system. The output of either trip system will trip both recirculation pumps motor breakers.
| |
| E1-32
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions 5.4 TS 3.3.5.1 - ECCS Instrumentation The ECCS instrumentation also employs diversity in the number and variety of different inputs which will actuate the associated equipment. The ECCS instrumentation actuates CS, LPCI, HPCI, ADS, and the emergency DGs The ECCS Instrumentation also employs diversity in the number and variety of different inputs which will actuate the associated equipment. Table E1-7 below presents the TS 3.3.5.1 logic descriptions for the functions listed in TS Table 3.3.5.1-1.
| |
| Table E1-7: ECCS Instrumentation Diversity Function Logic Logic Description Core Spray (CS) System Reactor water level is monitored by four redundant differential pressure instruments, which are, in turn, Function 1.a - Reactor connected to four trip units. The outputs of these Vessel Water LevelLow 2/4 trip units are connected to relays whose contacts Low are arranged in a one-out-of-two taken twice logic (i.e., two trip systems) to initiate CS.
| |
| Drywell pressure is monitored by four redundant pressure switches, which are, in turn, connected to Function 1.b - Drywell four trip units. The outputs of these trip units are 2/4 PressureHigh connected to relays whose contacts are arranged in a one-out-of-two taken twice logic (i.e., two trip systems) to initiate CS.
| |
| The Reactor Steam Dome Pressure - Low Function 1.c - Reactor (permissive) variable is monitored by two pressure Steam Dome Pressure 1/2 switches connected to relays arranged in a one-out-Low (Permissive) of-two logic to provide permissive for opening injection valve of low pressure ECCS subsystems.
| |
| The CS pump discharge flow is monitored by a flow switch (one per each of the two CS pumps). When the pump is running and discharge flow is low Function 1.d - Core Spray enough so that pump overheating may occur, the Pump Discharge Flow 1/1 minimum flow return line valve is opened. The valve Low (Bypass) is automatically closed if flow is above the minimum flow setpoint to allow the full system flow assumed in the accident analysis.
| |
| Function 1.e - Core Spray Pump Start-Time Delay 1/1 There is one time delay relay per CS pump.
| |
| Relay Low Pressure Coolant Injection (LPCI) System Function 2.a - Reactor Vessel Water LevelLow 2/4 See Function 1.a above (same logic for LPCI).
| |
| Low Function 2.b - Drywell 2/4 See Function 1.b above (same logic for LPCI).
| |
| PressureHigh E1-33
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-7: ECCS Instrumentation Diversity Function Logic Logic Description Function 2.c - Reactor Steam Dome Pressure 1/2 See Function 1.c above (same logic for LPCI).
| |
| Low (Permissive)
| |
| The Reactor Steam Dome variable is monitored by four pressure switches, which are, in turn, connected to multiple relays whose contacts are Function 2.d - Reactor arranged in a one-out-of-two taken twice logic.
| |
| Steam Dome Pressure 2/4 These instruments function to initiate closure of the Low (Break Detection) recirculation pump discharge valves to ensure that LPCI flow does not bypass the core when it injects into the recirculation lines.
| |
| Function 2.e - Low Pressure Coolant See Function 1.e. There is one time delay relay per Injection Pump Start-Time 1/1 LPCI pump loop.
| |
| Delay Relay Pumps B and D
| |
| Function 2.f - Low Pressure Coolant See Function 1.d above (same logic for LPCI; 1 flow 1/1 Injection Pump Discharge switch for each of the two RHR loops).
| |
| FlowLow (Bypass)
| |
| Recirculation Pump Differential Pressure-High (Break Detection) signals are initiated from eight Function 2.g -
| |
| differential pressure switches, four of which sense Recirculation Pump 4/8 the pressure differential between the suction and Differential Pressure-High discharge of each of the two recirculation pumps.
| |
| (Break Detection)
| |
| The relay outputs are arranged in a one-out-of-two taken twice logic for each pump.
| |
| Recirculation Riser Differential PressureHigh (Break Detection) signals are initiated from four differential pressure switches that sense the pressure differential between the A recirculation loop riser and the B recirculation loop riser. If, after Function 2.h -
| |
| a small time delay, the pressure in loop A is not Recirculation Riser 2/4 indicating higher than loop B pressure, the logic will Differential Pressure select the B loop for injection. If recirculation loop A High (Break Detection) pressure is indicating higher than loop B pressure, the logic will select the A loop for LPCI injection.
| |
| The Recirculation Riser Differential PressureHigh (Break Detection) output signals are combined in a one-out-of-two taken twice logic E1-34
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-7: ECCS Instrumentation Diversity Function Logic Logic Description Function 2.i -
| |
| Recirculation Pump Differential Pressure Time Recirculation Pump DelayRelay (Break Detection) signals are initiated Differential Pressure Time 2/2 by two time delay relays; one time delay relay per DelayRelay (Break each of the two loops.
| |
| Detection)
| |
| Function 2.j - Reactor Reactor Steam Dome Pressure Time DelayRelay Steam Dome Pressure 2/2 (Break Detection) signals are initiated from two time Time Delay Relay delay relays; one for each loop.
| |
| (Break Detection)
| |
| Function 2.k -
| |
| Recirculation Riser Recirculation Riser Differential Pressure Time Differential Pressure Time 2/2 DelayRelay (Break Detection) signals are initiated DelayRelay (Break by two time delay relays; one for each loop.
| |
| Detection)
| |
| High Pressure Coolant Injection (HPCI) System Four independent transmitters are connected to Function 3.a - Reactor multiple trip units. The outputs of the trip units are Vessel Water Level -Low 2/4 connected to relays whose contacts are arranged in Low a one-out-of-two taken twice logic to initiate HPCI.
| |
| Four independent pressure switches are connected Function 3.b - Drywell 2/4 to relays whose contacts are arranged in a one-out-Pressure High of-two taken twice logic to initiate HPCI.
| |
| Reactor Vessel Water Level - High signals for HPCI Function 3.c - Reactor are initiated from two differential pressure Vessel Water Level 2/2 instruments from the narrow range water level High measurement instrumentation. Both signals are required in order to close the HPCI injection valve.
| |
| Contaminated Condensate Storage Tank Level Low signals are initiated from four level switches Function 3.d - (two associated with each CCST). The output from 1/2 Contaminated these switches are provided to the logic of HPCI (per CCST)
| |
| Condensate Storage Tank system. The logic is arranged such that any level (CCST) Level Low switch can cause the suppression pool suction valves to open and the CCST suction valve of both units to close.
| |
| Suppression Pool Water LevelHigh signals are Function 3.e - initiated from two level switches. The logic is Suppression Pool Water 1/2 arranged such that either switch can cause the LevelHigh suppression pool suction valves to open and the CCST suction valve to close.
| |
| E1-35
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-7: ECCS Instrumentation Diversity Function Logic Logic Description One differential pressure switch is used to detect the Function 3.f - High HPCI System flow rate. The logic is arranged such Pressure Coolant 1/1 that the switch causes the minimum flow valve to Injection Pump Discharge open. The logic will close the minimum flow valve Flow Low (Bypass) once the closure setpoint is exceeded.
| |
| Function 3.g - Manual There is one manual initiation push button for the 1/1 Initiation HPCI system.
| |
| Automatic Depressurization System (ADS) Trip Systems A & B Reactor Vessel Water LevelLow Low signals are initiated from four differential pressure instruments.
| |
| Two channels input to ADS trip system A, while the other two channels input to ADS trip System B.
| |
| The ADS logic (low low reactor vessel and high drywell pressure) in each trip system is arranged in two strings. Each string has a contact from a Function 4.a/5.a - Reactor Vessel Water Level-Low Low and Drywell Reactor Vessel Water 2/4 PressureHigh function channel. In addition, each LevelLow Low string receives a contact input of a pressure switch associated with each CS and LPCI pump via the use of auxiliary relays and one string includes the ADS initiation timer. All contacts in both logic strings must close, the ADS initiation timer must time out, and a CS or LPCI pump discharge pressure signal must be present to initiate an ADS trip system. Either the A or B trip system will cause all the ADS relief valves to open.
| |
| Drywell PressureHigh signals are initiated from Function 4.b/5.b - Drywell four pressure switches that sense drywell pressure 2/4 PressureHigh (two for each trip system). See Function 4.a/5.a for logic description.
| |
| Function 4.c/5.c - There are two Automatic Depressurization System Automatic Initiation Timer relays, one in each of the two ADS 1/2 Depressurization System trip systems. See Function 4.a/5.a for logic Initiation Timer description.
| |
| E1-36
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-7: ECCS Instrumentation Diversity Function Logic Logic Description Two pressure switches (twelve total) on the discharge of each core spray and each LPCI pump are connected through relays in redundant groups so that each ADS trip system is blocked from Function 4.d/5.d - Core actuating unless at least one low pressure pump Spray Pump Discharge 2/4 shows verified discharge pressure. In order to Pressure-High generate an ADS permissive in one trip system, it is necessary that only one pump (both channels for the pump) indicate the high discharge pressure condition. See Function 4.a/5.a for logic description.
| |
| Function 4.e/5.e - Low Pressure Coolant 2/8 See Function 4.d/5.d above for logic description.
| |
| Injection Pump Discharge Pressure-High Function 4.f/5.f - There are two Automatic Depressurization System Automatic Low Low Water Level Actuation Timer relays, one in Depressurization System 1/2 each of the two ADS trip systems. Either the A or B Low Low Water Level trip system will cause all the ADS relief valves to Actuation Timer open.
| |
| 5.5 TS 3.3.5.3 - RCIC System Instrumentation The RCIC System instrumentation initiates actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink (the main condenser) and normal coolant makeup flow from the Reactor Feedwater System is unavailable. RCIC Instrumentation employs diversity in the number of different inputs which will actuate the associated equipment.
| |
| Table E1-8 below presents the TS 3.3.5.3 logic descriptions for the functions listed in TS Table 3.3.5.3-1.
| |
| Table E1-8: RCIC System Instrumentation Diversity Function Logic Logic Description Four transmitters are connected to four trip units.
| |
| Function 1 - Reactor The outputs of the trip units are connected to relays Vessel Water LevelLow 2/4 whose contacts are arranged in a one-out-of-two Low taken twice logic arrangement.
| |
| Function 2 - Reactor The Reactor Vessel Water Level - High trip is 2/2 Vessel Water LevelHigh arranged in a two-out-of-two logic.
| |
| Contaminated Condensate Storage Tank Level Function 3 -- Low signals are initiated from four level switches Contaminated 1/2 (two associated with each CCST). The output from Condensate Storage Tank (per CCST) these switches are provided to the logics of both (CCST) Level-Low HPCI Systems. The logic is arranged such that any level switch can cause the suppression pool suction E1-37
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-8: RCIC System Instrumentation Diversity Function Logic Logic Description valves to open and the CCST suction valve of both units to close.
| |
| Function 4 - Suppression Suppression pool water level signals are initiated 1/2 Pool Water LevelHigh from two level switches in a one-out-of-two logic Function 5 - Manual There is one manual initiation push button for the 1/1 Initiation RCIC system.
| |
| 5.6 TS 3.3.6.1 - Primary Containment Isolation Instrumentation Primary containment isolation instrumentation also employs diversity in the number and variety of different inputs which will actuate the associated equipment. The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of primary containment and reactor coolant pressure boundary (RCPB) isolation. Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a primary containment isolation signal to the isolation logic. Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation logics are (a) reactor vessel water level, (b) area ambient temperatures, (c) main steam line (MSL) flow measurement, (d) SLC system initiation, (e) main steam line pressure, (f) HPCI and RCIC steam line flow, (g) drywell radiation and pressure, (h) HPCI and RCIC steam line pressure, and (i) reactor vessel pressure. Redundant sensor input signals from each parameter are provided for initiation of isolation. The only exception is SLC system initiation. Primary containment isolation instrumentation has inputs to the trip logic of the isolation functions listed below. Table E1-9 below presents the TS 3.3.6.1 logic descriptions for all the functions listed in TS Table 3.3.6.1-1.
| |
| Table E1-9: Primary Containment Isolation Instrumentation Diversity Function Logic Logic Description Main Steam Line Isolation The Reactor Vessel Water LevelLow Low, the Main Steam Line PressureLow, and the Main Steam Line PressureTimer Functions receive inputs from four channels. One channel associated with each Function inputs to one of four trip strings.
| |
| Function 1.a - Reactor Two trip strings make up a trip system and both trip Vessel Water LevelLow 2/4 systems must trip to cause an isolation of all main Low steam isolation valves (MSIVs), MSL drain valves, and recirculation loop sample isolation valves. Any channel will trip the associated trip string. Only one trip string must trip to trip the associated trip system.
| |
| The trip strings are arranged in a one-out-of-two taken twice logic to initiate isolation.
| |
| E1-38
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-9: Primary Containment Isolation Instrumentation Diversity Function Logic Logic Description Function 1.b - Main Steam Line Pressure 2/4 See Function 1.a above.
| |
| Low Function 1.c - Main Steam Line Pressure 2/4 See Function 1.a above.
| |
| Timer The Main Steam Line Flow - High and Main Steam Tunnel Temperature - High Functions each contain 16 channels. Each PCIS trip channel receives four inputs from each of these functions, one flow input Function 1.d - Main 2/4 from each MSL and one temperature input from Steam Line Flow-High each of the four areas monitored. Any one of these inputs will trip the associated PCIS trip channel.
| |
| The four PCIS trip channels are arranged in a one-out-of-two taken twice logic.
| |
| Function 1.e - Main Steam Line Tunnel 2/4 See Function 1.d above.
| |
| TemperatureHigh Primary Containment Isolation The Reactor Vessel Water LevelLow and Drywell Pressure-High Functions receive inputs from four channels. One channel associated with each Function inputs to one of four trip strings. Two trip strings make up a trip system and both trip systems Function 2.a - Reactor must trip to cause an isolation of the PCIVs 2/4 Vessel Water LevelLow identified in Reference 1 of TS B 3.3.6.1 (Reference [4]). Any channel will trip the associated trip string. Only one trip string must trip to trip the associated trip system. The trip strings are arranged in a one-out-of-two taken twice logic to initiate isolation.
| |
| Function 2.b - Drywell 2/4 See Function 2.a above.
| |
| PressureHigh The Drywell RadiationHigh Function receives input from two radiation detector assemblies each connected to a switch. Each switch actuates two contacts. Each contact inputs to one of four trip strings. Two trip strings make up a trip system and Function 2.c - Drywell 2/4 both trip systems must trip to cause an isolation of RadiationHigh the PCIVs identified in Reference 1 of TS B 3.3.6.1 (Reference [4]). The contacts associated with the same switch provide input to both trip strings in the same trip system. Any contact will trip the associated trip string. The trip strings are arranged E1-39
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-9: Primary Containment Isolation Instrumentation Diversity Function Logic Logic Description in a one-out-of-two taken twice logic. A channel is considered to include a radiation detector assembly, a switch, and one of two contacts.
| |
| HPCI System Isolation The HPCI Steam FlowHigh and HPCI Steam Flow 2/2 Timer Functions each receive input from two Function 3.a - HPCI (for both channels, with each channel in one trip system Steam Line Flow-High isolation using a one-out-of-one logic. Each of the two trip valves) systems is connected to one of the two valves on the HPCI Steam supply penetration.
| |
| Function 3.b - HPCI 2/2 See Function 3.a above.
| |
| Steam Line Flow-Timer The HPCI and RCIC Steam Supply Line Pressure Low Functions receive inputs from four steam 4/4 Function 3.c - HPCI supply pressure channels for each system. The (for both Steam Supply Line outputs from HPCI steam supply pressure channels isolation PressureLow are each connected to two two-out-of-two trip valves) systems. Each trip system isolates one valve on the HPCI steam supply penetration.
| |
| The HPCI Drywell PressureHigh Function receives input from four channels. Two channels provide input to one trip system and the other two channels provide input to a second trip system. In addition, four HPCI Steam Supply Line Pressure 4/4 Low Function channels are also connected to these Function 3.d - Drywell (for both trip systems. Each of the two trip systems receives PressureHigh isolation input from two additional HPCI Steam Supply Line valves) PressureLow Function channels. Each trip system is arranged such that one channel associated with each Function must trip in order to initiate isolation of one HPCI vacuum breaker isolation valve. The logic in each trip system is one-out-of-two for each Function.
| |
| The HPCI Turbine Area TemperatureHigh Function receives input from four channels. Two channels monitor the area near the steam supply 4/4 line while the other two channels monitor the Function 3.e - HPCI (for both temperature near the turbine exhaust rupture disc.
| |
| Turbine Area HPCI Each of the two trip systems receives input from one TemperatureHigh valves) channel in each of the two areas. Each trip system is arranged such that both channels must trip in order to initiate isolation. This is effectively a two-out-of-two logic arrangement. Each of the two trip E1-40
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-9: Primary Containment Isolation Instrumentation Diversity Function Logic Logic Description systems is connected to one of the two valves on the HPCI steam supply penetration.
| |
| RCIC System Isolation The RCIC Steam FlowHigh and RCIC Steam FlowTimer Functions each receive input from two channels. Each channel is connected to two trip Function 4.a - RCIC 1/2 systems, each using a one-out-of-two logic. Each of Steam Line Flow-High the two trip systems is connected to both RCIC steam supply isolation valves, such that any trip system will isolate both valves.
| |
| Function 4.b - RCIC 1/2 See Function 4.a above.
| |
| Steam Line Flow-Timer The HPCI and RCIC Steam Supply Line Pressure Low Functions receive inputs from four steam Function 4.c - RCIC supply pressure channels for each system. The Steam Supply Line 2/4 RCIC Steam Supply Line PressureLow channels PressureLow are arranged in a one one-out-of-two twice trip system. The trip system is connected to both RCIC steam supply isolation valves.
| |
| The RCIC Turbine Area TemperatureHigh Function receives input from four channels. The four channels monitor the area near the RCIC turbine. Each of the two trip systems receives input Function 4.d - RCIC from the four channels. Each trip system is Turbine Area 2/4 arranged in a one-out-of-two taken twice logic to TemperatureHigh initiate isolation. Each of the two trip systems is connected to both RCIC steam supply isolation valves, such that any trip system will isolate both valves.
| |
| Reactor Water Cleanup System Isolation The SLC System Initiation Function receives input from the SLC initiation switch. The switch provides trip signal inputs to both trip systems in any position other than "OFF". The other switch positions are Function 5.a - SLC 1/1 SYS 1, SYS 2, SYS 1+2 and SYS 2+1. For the System Initiation purpose of this Specification, the SLC initiation switch is considered to provide 1 channel input into each trip system. Each of the two trip systems is connected to one of the two RWCU valves.
| |
| The Reactor Vessel Water LevelLow Isolation Function receives input from four reactor vessel Function 5.b - Reactor 2/4 water level channels. Each channel inputs into one Vessel Water LevelLow of four trip strings. Two trip strings make up a trip system and both trip systems must trip to cause an E1-41
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-9: Primary Containment Isolation Instrumentation Diversity Function Logic Logic Description isolation of the reactor water cleanup (RWCU) valves. Any channel will trip the associated trip string. Only one trip string must trip to trip the associated trip system. The trip strings are arranged in a one-out-of-two taken twice logic to initiate isolation.
| |
| RHR Shutdown Cooling System Isolation The Reactor Vessel Pressure-High Function receives input from two channels, both of which 2/2 provide input to both trip systems. Any channel will Function 6.a - Reactor (for both trip both trip systems. This is a one-out-of-two logic Vessel PressureHigh valves) for each trip system. Each of the two trip systems is connected to one of the two valves on the RHR SDC suction penetration.
| |
| The Reactor Vessel Water LevelLow function receives input from four reactor vessel water level channels. Each channel inputs into one of four trip strings. Two trip strings make up a trip system and Function 6.b - Reactor both trip systems must trip to cause an isolation of 2/4 Vessel Water LevelLow the RHR SDC suction isolation valves. Any channel will trip the associated trip string. Only one trip string must trip to trip the associated trip system.
| |
| The trip strings are arranged in a one-out-of-two taken twice logic to initiate isolation.
| |
| 5.7 TS 3.3.8.1 - LOP Instrumentation Each 4160 V Essential Service System (ESS) bus has its own independent LOP instrumentation and associated trip logic. The voltage for each bus is monitored at two levels, which can be considered as two different undervoltage Functions: Loss of Voltage and Degraded Voltage. Each Division 1 and 2 4160 V ESS Bus Loss of Voltage and Degraded Voltage.
| |
| Each Division 1 and 2 4160 V ESS Bus Loss of Voltage and Degraded Voltage Function is monitored by two undervoltage relays for each ESS bus, whose outputs are arranged in a two-out-of-two logic configuration (Reference [5]). When, on decreasing voltage, the 4160 V ESS Bus Undervoltage (Loss of Voltage) Function setpoint has been exceeded on both relay channels, the Loss of Voltage Function sends a LOP signal to the respective bus load shedding scheme and starts the associated DG. For the Degraded Voltage Function, one Bus Undervoltage/Time Delay Function (two channels) and one Time Delay Function (one channel) are included. The Time Delay Function associated with the Bus Undervoltage relay is inherent to the Bus Undervoltage - Degraded Voltage relay and is nominally adjusted to seven seconds to prevent circuit initiation caused by grid disturbances and motor starting transients. The Bus Undervoltage/Time Delay Function provides input to the Time Delay Function. The Time Delay E1-42
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Function relay is nominally adjusted to five minutes to allow time for the operator to attempt to restore normal bus voltage.
| |
| Table E1-10 below presents the TS 3.3.8.1 logic descriptions for all the functions listed in TS Table 3.3.8.1-1.
| |
| Table E1-10: LOP Instrumentation Diversity Function Logic Logic Description 4160 V Essential Service System Bus Undervoltage (Loss of Voltage)
| |
| The Loss of Voltage Function is monitored by two undervoltage relay pairs for each Division 4160 V ESS Bus, where outputs are arranged in a two-out-2/2 of-two logic configuration. When, on decreasing Function 1 - Loss of (for each voltage, the 4160 V ESS Bus Undervoltage (Loss of Voltage 4kV ESS Voltage) Function setpoint has been exceeded on bus) both relay channels, the Loss of Voltage Function sends a LOP signal to the respective bus load shedding scheme and starts the associated DG.
| |
| 4160 V Essential Service System Bus Undervoltage (Degraded Voltage)
| |
| Each Division 1 and 2 4160 V ESS Bus Loss of Voltage and Degraded Voltage Function is monitored by two undervoltage relays for each ESS bus, whose outputs are arranged in a two-out-of-two logic configuration. For the Degraded Voltage Function, one Bus Undervoltage/Time Delay Function (two channels) and one Time Delay Function (one channel) are included. The Bus Undervoltage/Time Delay Function provides input to the Time Delay Function. The Time Delay Function relay is nominally adjusted to five minutes to allow 2/2 time for the operator to attempt to restore normal Function 2.a - Bus (for each bus voltage. When a Bus Undervoltage/Time Delay Undervoltage/Time Delay 4kV ESS Function setpoint has been exceeded and persists bus) for seven seconds on both relay channels, a control room annunciator alerts the operator of the degraded voltage condition, and the five minute Time Delay Function timer is initiated. If the degraded voltage condition does not clear within five minutes, the five minute Time Delay Function relay sends a LOP signal to the respective bus load shedding scheme and starts the associated DG. If a degraded voltage condition exists coincident with an ECCS actuation signal, the five minute Time Delay Function is bypassed such that load shedding and the associated DG start will be initiated following the E1-43
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table E1-10: LOP Instrumentation Diversity Function Logic Logic Description seven second time delay (Bus Undervoltage/Time Delay Function).
| |
| 1/1 Function 2.b - Time Delay (for each See Function 2.a logic description above.
| |
| (No LOCA) 4kV ESS bus)
| |
| : 6. Regulatory Guide 1.174, Revision 3, Section 2.1.1 - Defense-in-Depth In accordance with the principles contained within Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, defense-in-depth consists of several elements and consistency with the defense-in-depth philosophy is maintained if the following occurs:
| |
| : 1) Preserve a reasonable balance among the layers of defense.
| |
| Current TS reflect this balance by allowing one sensor module or channel to be placed in trip, while preserving the fundamental safety function of the applicable system. Tripping an inoperable channel does not affect the number of channels required to provide the safety function. Even in the TS condition for two channels in a function inoperable, the fundamental safety function is preserved since sufficient operable channels remain in the function.
| |
| : 2) Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.
| |
| No programmatic activities are relied upon as compensatory measures when one or two channels of the applicable instrumentation are inoperable. The remaining operable channels for that function are fully capable of performing the safety function of the applicable system.
| |
| : 3) Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.
| |
| System redundancy, independence and diversity remain the same as in the as-designed condition. The number of operable functions has not been decreased (diversity), the number of minimum operable channels to perform the safety function has not been decreased, and the channels remain independent as originally designed, even with one channel inoperable.
| |
| E1-44
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions
| |
| : 4) Preserve adequate defense against potential common cause failures.
| |
| Implementation of a RICT Program does not impact the original determination of common-cause failure for the applicable instrumentation and its functions. It may allow the allowed outage time to be extended for one or two channels in a function to be inoperable prior to placing the channel in trip. Placing the channel in trip fulfils one of the two required channels in trip needed to perform the safety function.
| |
| : 5) Maintain multiple fission product barriers.
| |
| Fission product barriers are not affected by this LAR request.
| |
| : 6) Preserve sufficient defense against human errors Proposed changes include revisions to operational procedures and training to Operations and Engineering personnel to prevent occurrence of human errors.
| |
| : 7) Continue to meet intent of the plant's design criteria.
| |
| Implementation of the RICT Program does not change the design, configuration, or method of operation of the plant. Except for the local intense precipitation (LIP) barrier modifications discussed in Enclosure 4, Section 5.2.3.3, the proposed changes do not involve a physical alteration of the plant (i.e., no new or different kind of equipment will be installed).
| |
| : 7. Description of Electrical System Capabilities at QCNPS The onsite AC power system consists of two main generators, two main step-up transformers, two unit auxiliary transformers (UATs), two reserve auxiliary transformers (RATs), distribution buses, three standby emergency diesel generators (DGs), and two standby station blackout (SBO) DGs. The AC distribution system has nominal ratings of 13.8 kV, 4160 V, 480/277 V, and 208/120 V.
| |
| 7.1 AC Sources The unit Class 1E AC Electrical Power Distribution System AC sources consist of the offsite power sources, and the onsite standby power sources (DGs 1, 2, and 1/2). The design of the AC electrical power system provides independence and redundancy to ensure an available source of power to the Engineered Safety Feature (ESF) systems.
| |
| The Class 1E unit AC distribution system is, for the most part, divided into redundant load groups (Divisions 1 and 2), so loss of any one group does not prevent the minimum safety functions from being performed. The exception is that the opposite unit's AC Electrical Power Distribution System powers shared loads (i.e., standby gas treatment subsystem, Control Room Emergency Ventilation (CREV) System (Unit 2 only), and Control Room Emergency Ventilation Air Conditioning (AC) System (Unit 2 only)). Although shared by both units, the CREV System and Control Room Emergency Ventilation AC System are single train systems that are powered E1-45
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions only from a single Unit 1 motor control center. Each unit's load group has connections to two physically independent offsite power sources and a single DG.
| |
| Offsite power is supplied to the 345 kV switchyard from the transmission network by five transmission lines. From the 345 kV switchyard, one qualified electrically and physically separated circuit normally provides AC power, through RAT 12, to 4160 V ESS bus 13-1 via ESS bus 13 to supply the Division 1 loads of Unit 1. From the same switchyard, another qualified, electrically and physically separated circuit normally provides AC power through RAT 22 to 4160 V ESS bus 23-1 via ESS bus 23 to supply the Division 1 loads of Unit 2. UAT 11, which is normally supplied by the Unit 1 main generator, is normally aligned to supply the Unit 1 Division 2 4160 V ESS bus 14-1 via ESS bus 14. Finally, UAT 21, which is normally supplied by the Unit 2 main generator, is normally aligned to supply the Unit 2 Division 2 4160 V ESS bus 24-1 via ESS bus 24.
| |
| When a main generator is not operating, the loads fed from the UAT are automatically transferred to the RAT on a generator trip (RAT 12 will supply 4160 V ESS bus 14-1 via 4160 V ESS bus 14 and RAT 22 will supply 4160 V ESS bus 24-1 via 4160 V ESS bus 24). The given unit's RAT is the primary (normal) offsite source to the Division 1 and 2 load groups. The RAT of the opposite unit provides the second (alternate) qualified offsite source through bus ties provided between the corresponding ESS buses of the two units. Additionally, the UAT of either unit provides another source of offsite power to the ESS buses only when the unit is shutdown and the UAT is being backfed from the grid. Physical changes to the generator links are required to place the unit in an alignment to allow backfeed. The offsite AC electrical power sources are designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions.
| |
| The UAT must be physically connected, breakers closed, to the 4 kV bus to be considered an offsite source while in the backfeed condition. This is due to an interlock that will prevent the manual closing of the UAT breakers to the 4 kV bus with a LOCA signal present.
| |
| RATs 12 and 22 are sized to accommodate the simultaneous starting of all ESF loads on receipt of an accident signal without the need for load sequencing. The onsite standby power source for 4160 V ESS buses 13-1, 14-1, 23-1, and 24-1 consists of three DGs. DGs 1 and 2 are dedicated to ESS buses 14-1 and 24-1, respectively. DG 1/2 is a shared power source and can supply either Unit 1 ESS bus 13-1 or Unit 2 ESS bus 23-1. A DG starts automatically on a loss of coolant accident (LOCA) signal (i.e., low reactor water level signal or high drywell pressure signal) or on an ESS bus degraded voltage or undervoltage signal. After the DG has started, it automatically ties to its respective bus after offsite power is tripped and certain permissives are met as a consequence of ESS bus undervoltage or degraded voltage, independent of or coincident with a LOCA signal. The DGs also start and operate in the standby mode without tying to the ESS bus on a LOCA signal alone. In the event of a LOCA on a unit, DG 1/2 will start and supply the unit (bus 13-1 or 23-1) experiencing the accident if no offsite power is available. This is accomplished by using the accident signal to prevent the DG 1/2 output breaker from closing on the non-accident unit. Following the trip of offsite power, buses 13-1, 14-1, 23-1, and 24-1 are automatically disconnected from their normal supply and all nonessential loads are disconnected from the ESS bus except the 480 V ESS bus. When the DG is tied to the ESS bus, loads are then sequentially connected to its respective ESS bus, if a E1-46
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions LOCA signal is present, by the sequencing logic. The sequencing logic controls the starting signals to motor breakers to prevent overloading the DG.
| |
| In the event of a loss of offsite power, the ESF electrical loads are automatically connected to the DGs in sufficient time to provide for safe reactor shutdown and to mitigate the consequences of a Design Basis Accident (DBA) such as a LOCA.
| |
| Certain required plant loads are returned to service in a predetermined sequence to prevent overloading of the DGs in the process. Within ~40 seconds after the initiating signal is received, all automatic and permanently connected loads needed to recover the unit or maintain it in a safe condition are returned to service.
| |
| See included UFSAR Figure 8.3-1 below for high level overview of the Emergency Power System (Reference [4]).
| |
| E1-47
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions l 4!1 KV $W'tD .llti1I...l .IIWL.J l4!1 KV SW'l'D RCS AUX vJ.._, vJ.._, UNO NIX UNIT MJlC vJ.._, vJ.._, ACS AUX TIWIS 22 JR.fljS2t TlWIS II TRANS 12 1:lel 13" 14 12 NC N O 4ICV ~NC 1l 1312 NC
| |
| ~
| |
| ,.,~ 0
| |
| ~ ,.,
| |
| ~
| |
| C')
| |
| C
| |
| ~"Tl c:~
| |
| o- z zC'> Q ~o
| |
| .u,~ ,.,
| |
| C:
| |
| "D vl3
| |
| -Gl C...(>>
| |
| c:
| |
| zt.<
| |
| * ,.,i R--~
| |
| ,., !. ~
| |
| VI "'~
| |
| ~I:
| |
| ID 0 ID z ID Figure E1-1: UFSAR Figure 8.3-1 Emergency Power System E1-48
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions 7.2 Distribution Systems The onsite Class 1E AC electrical power distribution system for each unit is divided into redundant and independent AC electrical power distribution subsystems.
| |
| Each AC distribution subsystem consists of two 4160 V ESS buses having an offsite source of power as well as an onsite DG source. During normal operation, each subsystem's ESS buses are connected such that power is supplied to the Division 2 4160 V loads from the unit's main generator through a UAT and from the 345 kV system through the RAT to supply the Division 1 4160 V loads. The UAT and RAT are connected in a normal alternate power source arrangement for each of the 4160 V divisions (i.e., the RAT provides alternate power for the Division 2 ESS buses and the UAT for the Division 1 ESS buses). During a loss of the normal offsite power source to the 4160 V ESS buses, the alternate supply breaker attempts to close.
| |
| If all offsite sources are unavailable, the onsite emergency unit DGs supply power to the 4160 V ESS buses. With a LOCA signal present, the UAT feed to the ESS buses will not attempt to close on a loss of normal power from the RAT. This is to prevent multiple starts of equipment in the case of a LOCA coincident with an open phase event on the offsite source to the RAT.
| |
| Each AC distribution subsystem also includes 480 VAC ESS buses 18 and 19 (Unit 1) and buses 28 and 29 (Unit 2), associated motor control centers, transformers and distribution panels.
| |
| The 120 VAC instrument bus is normally powered from 480 VAC bus 18-2 for Unit 1 and 480 VAC MCC 28-2 for Unit 2. The alternate power supply for the Unit 1 120 VAC instrument bus is supplied from 480 VAC MCC 15-2 and the Unit 2 120 VAC instrument bus is supplied from 480 VAC MCC 25-2. On a loss of normal power to the instrument bus an automatic bus transfer (ABT) switches to the alternate supply and automatically switches back to the normal supply when the normal supply is restored. However, the instrument bus ABT is only provided for reliability and is not required to be operable (i.e., only one power source to the instrument bus is required).
| |
| The 120 VAC essential services bus is supplied by a static uninterruptible power supply (UPS) or an alternate source from 18-2 (28-2). Power to the UPS is supplied, in order of preference; for Unit 1 by 480 VAC bus 18, 250 VDC MCC 1, or 480 VAC bus 17; and for Unit 2 by 480 VAC bus 28, 250 VDC MCC 2, or 480 VAC bus 26.
| |
| There are two independent 250 VDC station service electrical power distribution subsystems and two independent 125 VDC electrical power distribution subsystems that support the necessary power for ESF functions. The 250 VDC electrical power distribution subsystem provides motive power to large DC loads such as DC motor-driven pumps and valves.
| |
| Division 1 and 2 125 VDC electrical power distribution subsystems provide control power to selected safety related equipment as well as circuit breaker control power for 4160 V, 480 V, control relays, and annunciators. The Division 2 125 VDC subsystem for each unit is provided power by the opposite unit's battery and provides control power to a shared standby gas treatment subsystem.
| |
| E1-49
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions The required electrical power distribution subsystems listed in Table B 3.8.7-1 for Unit 1 and Table B 3.8.7-2 for Unit 2 (provided below) ensure the availability of AC and DC electrical power for the systems required to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence (AOO) or a postulated DBA.
| |
| Table B 3. B. 7-1 (page 1 of Unit 1 AC ad DC Electr i cal Powe Di str i but i on Systems TYPE VOLTAGE DIVISIO }(l)(Dl DIVIS ION 2<*Jlbl AC safety bus 4 60 V ESS buses 13, 13-1 ESS b ses 14 , 14-1 480 V ESS bus 18 ESS b s 19 120 V Uni essentia l A services bus , unit ins rument bus 250 voe buses 250 V NA TB MCC 1, RB CC lA ,
| |
| RB MCC lB 125 voe buses 125 V TB main b ses IA , TB ain bus 2A ;
| |
| l A-1 ; TB reserve buses RB distrib ion lB' 1 and lB-1 pane l l Ca) ac h di vi s i on of t he AC and DC electrical power distribut i on systems i s a subsystem . The 250 voe buses constitute a s i ngle subsystem (Div i s i on 2) .
| |
| Cb) OPERABI LITY requ i rements of t he opposite unit's Division 1 and Di vision 2 AC and DC electrical power distribution systems require OPERABI LITY of t he 4160 VAC bus 24- 1, 480 VAC bus 29 , essential services 120 VAC bus , and 125 VDC bus 28 CDC Distribution Bus 28 is not a requ i ed di stribution bus when DC Bus 2B- l is energized by an OPE ABLE 125 VDC batte y (inc l uding one fu l l capacity charger) ,
| |
| prov i ded tat Bus 2B i s electrica l ly isolated f om Bus 2B-l vi a an open t i e brea er) . In the event Unit 2 i s shutdow in MOOE 4 or MOO E 5, bus 24-1 i s not a requ i red oppos i te unit 4160 VAC bus if bus 29 i s powered by bus 23- 1 via the 480 VAC b s 28/29 crosstie .
| |
| Cc) DC Di st 1but 1o Bus 18 i s not a req ired di strib tio bus hen DCB s lB -1 is ene gized by an OPERAB LE 125 VDC Battery ( i ncluding one full capacity c arge ), p ov i ded t hat Bus lB i s electrica l ly isolated f om Bus IB-1 via a open tie beaker .
| |
| Figure E1-2: TS Table B 3.8.7-1 Unit 1 AC and DC Electrical Power Distribution Systems E1-50
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions Table B 3.8 . 7-2 (page 1 of 1)
| |
| Unit 2 AC and DC Electrical Powe Di str i but i on Systems YPE VOLTAGE OIVISIO 1(1 bl OIVIS!Oll 2c,Ho>
| |
| AC safety bus 4160 V ESS buses 23 , 23-1 ESS bus 24, 24-1 480 V ESS bus 28 ESS bus 29 120 V Uni essen ia l A services b s , unit ins rument bus 250 voe buses 250 V A TB MCC 2 ,
| |
| RB MCC 2A ,
| |
| RB MCC 28 125 voe buses 125 V TB main bus 2A, TB main bus lA; 2A-l; TB reserve buses RB distrib tion 2s ec) , 28 - 1 pane l 2 Ca) ach di vi s i on oft e AC and DC electrical power distribut i on systems is a subsystem . The 250 voe buses constitute a single subsystem
| |
| <Div i s i on 2) .
| |
| {b) OP ABI LITY reQu i rements of t he opposite unit's Division 1 and
| |
| *vision 2 AC and DC electr*cal power dis ribution sys ems reQuire OPE RABI LITY oft e 4160 VAC bus 14-1, 480 VAC bus 19, essential services 120 VAC bus , and 125 voe bus 18 CDC Distrib ion Bus 18 is not a requ i ed di stribution bus whe DC Bus 18-1 is energized by an OPERABLE 125 VDC battery (inc l uding one fu l l capacity charger),
| |
| prov i ded tat Bus 18 is el ectrica l ly iso l ated from B s 18-1 vi a an open tie brea er) . In the event Unit 1 is shutdown in MODE 4 or ODE 5, bus 14-1 i s not a requ i red oppos i te unit 4160 VAC bus i f bus 19 i s powered by bus 13- 1 via the 480 VAC bus 18/19 crosstie .
| |
| {c) DC Di st i but i on Bus 2B i s no t a req ired distrib tio bus hen DCB s 2B - l is energized by an OPERABLE 125 voe Battery ( i ncluding one full capacity c arger) p ov i ded t hat Bus 28 is e l ec t rica l ly iso l ated from Bus 28-1 vi a an open tie breake Figure E1-3: TS Table B 3.8.7-2 Unit 2 AC and DC Electrical Power Distribution Systems
| |
| : 8. References
| |
| [1] Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines',"
| |
| dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| E1-51
| |
| | |
| License Amendment Request Enclosure 1 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 List of Revised Required Actions to Corresponding PRA Functions
| |
| [2] Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)
| |
| Guidelines," Revision 0, dated October 12, 2012 (ADAMS Accession No. ML12286A322)
| |
| [3] ASME Standard ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," dated February 2, 2009
| |
| [4] QCNPS Updated Final Safety Analysis Report (UFSAR), Revision 16, dated October 21, 2021
| |
| [5] QCNPS Technical Specification Bases, dated October 21, 2021 E1-52
| |
| | |
| ENCLOSURE 2 License Amendment Request Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Information Supporting Consistency with Regulatory Guide 1.200, Revision 2
| |
| | |
| License Amendment Request Enclosure 2 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2
| |
| : 1. Introduction This enclosure provides information on the technical adequacy of the Quad Cities Nuclear Power Station (QCNPS) Probabilistic Risk Assessment (PRA) internal events model (including flooding) and the QCNPS fire PRA (FPRA) model in support of the license amendment request to revise Technical Specifications to implement Nuclear Energy Institute (NEI) 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines" (Reference [14]).
| |
| Topical Report NEI 06-09, Revision 0-A, as clarified by the U.S. Nuclear Regulatory Commission (NRC) final safety evaluation of this report (Reference [15]), defines the technical attributes of a PRA model and its associated Configuration Risk Management Program (CRMP) tool, presently referred to as the Real-Time Risk (RTR) tool, required to implement this risk-informed application. Meeting these requirements satisfies Regulatory Guide (RG) 1.174 (Reference [16]) requirements for risk-informed plant-specific changes to a plant's licensing basis.
| |
| Constellation Energy Generation, LLC (CEG) employs a multi-faceted approach to establishing and maintaining the technical acceptability and fidelity of PRA models for all operating CEG nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update process and the use of self-assessments and independent peer reviews.
| |
| Section 2 outlines requirements related to the scope of the QCNPS PRA Full Power Internal Events (FPIE) model including internal flooding. Section 3 outlines the technical adequacy of the FPIE model and Section 4 describes the technical adequacy of the FPRA model used in this application.
| |
| : 2. Requirements Related to Scope of QCNPS PRA Models The PRA models discussed in this enclosure have been assessed against RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2 (Reference [1]) consistent with NRC Regulatory Information Summary (RIS) 2007-06 (Reference [2]).
| |
| Finding and Observation (F&O) closure reviews were conducted on the PRA models discussed in this enclosure. Closed findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, "Close-out of Facts and Observations (F&Os)" (Reference [3]) as accepted by NRC in the {{letter dated|date=May 3, 2017|text=letter dated May 3, 2017}} (Reference [4]).
| |
| Note that QCNPS FPIE PRA and FPRA models do not incorporate the risk impacts of external events. Enclosure 4 discusses the treatment of seismic risk and other external hazards for this application.
| |
| : 3. Scope and Technical Adequacy of QCNPS Internal Events and Internal Flooding PRA Model The QCNPS FPIE PRA model was peer reviewed in April 2017 using the NEI 05-04 process (Reference [5]), the PRA Standard (ASME/ANS RA-Sa-2009) [6] and RG 1.200, Revision 2.
| |
| E2-2
| |
| | |
| License Amendment Request Enclosure 2 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2 This Peer Review (Reference [7]) was a full-scope review of the technical elements of the Internal Events and Internal Flooding, at-power PRA. The FPIE PRA Peer Review team determined there were 31 unique finding level F&Os resulting in 14 Not-Met SRs. The findings from the Peer Review have been addressed in the Internal Events PRA model.
| |
| In February/March 2021, an F&O Closure Review (Reference [8]) was conducted for QCNPS.
| |
| The scope of the review included the Internal Events and Internal Flooding PRA model. The F&O Independent Assessment Team closed all Finding level F&Os. Currently, there are no open Finding level F&Os against the FPIE PRA model.
| |
| Given there are no partially resolved or open findings that may impact RICT calculations, the QCNPS FPIE PRA is of adequate technical capability to support the TSTF-505 program.
| |
| : 4. Scope and Technical Adequacy of QCNPS Fire PRA Model The QCNPS Fire PRA (FPRA) Peer Review (Reference [11]) was performed in June 2013 using the NEI 07-12 Fire PRA peer review process (Reference [12]), the ASME PRA Standard, ASME/ANS RA-Sa-2009 (Reference [6]), and RG 1.200, Revision 2 (Reference [1]). The purpose of this review was to establish the technical adequacy of the FPRA for the spectrum of potential risk-informed plant licensing applications for which the FPRA may be used. The FPRA Peer Review was a full-scope review of all technical elements of the QCNPS at-power FPRA against all technical elements in Part 4 of the ASME/ANS PRA Standard, including the referenced Internal Events Supporting Requirements (SRs) in Part 2. The Fire PRA Peer Review team determined there were 83 unique finding level F&Os resulting in 49 Not-Met SRs.
| |
| In February/March 2021, an F&O Closure by Independent Assessment was conducted for Quad Cities (Reference [8]). The scope of the review included all FPRA Peer Review finding level F&Os from the 2013 peer review; 76 of the 2013 peer review finding level F&Os were closed during this F&O closure review. Of the 49 SRs that were previously assessed as Not Met, 47 were assessed as Met to at least Capability Category II and two remained as Not Met.
| |
| A FPRA focused scope peer review (FSPR) was conducted in February 2021 [9]. This FSPR resulted in an additional five finding level F&Os. All reviewed SRs were Met to at least a Capability Category II.
| |
| In May 2021, a follow-on F&O closure review was conducted to assess the remaining seven finding level F&Os from the February 2021 F&O closure review and the five Finding level F&Os from the February 2021 FSPR (Reference [10]). All 12 finding level F&Os were assessed as closed during this review and all reviewed SRs were Met to at least a Capability Category II (including the two SRs that previously remained Not Met). Therefore, all previous finding level F&Os were closed and all applicable SRs were assessed as Met to at least Capability Category II.
| |
| Finally, another FSPR was conducted, also in May 2021 (Reference [13]). This FSPR resulted in one documentation finding level F&O (F&O 9-1) and one documentation SR assessed as Not Met (SR FQ-F1). The F&O identified that detailed descriptions of significant contributors including cutsets, accident sequences, etc. were not provided. This F&O is addressed by and resolved in the FPRA documentation prepared for the model used in support of this license E2-3
| |
| | |
| License Amendment Request Enclosure 2 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2 amendment request (LAR) (however, a formal F&O Closure review has not been completed for this remaining finding). Detailed descriptions of significant cutsets, accident sequences, etc.
| |
| were added to the documentation.
| |
| Given there are no partially resolved or open technical findings that may impact RICT calculations, and the only remaining Finding is related to documentation, the QCNPS FPRA is of acceptable technical capability to support the TSTF-505 program.
| |
| : 5. References
| |
| [1] Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, U.S.
| |
| Nuclear Regulatory Commission, dated March 2009 (ADAMS Accession No. ML090410014)
| |
| [2] NRC Regulatory Issue Summary 2007-06, "Regulatory Guide 1.200 Implementation,"
| |
| dated March 22, 2007 (ADAMS Accession No. ML070650428)
| |
| [3] NEI Letter to NRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," dated February 21, 2017 (ADAMS Accession No. ML17086A431)
| |
| [4] NRC Letter to Mr. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os)," dated May 3, 2017 (ADAMS Accession No. ML17079A427)
| |
| [5] NEI 05-04, "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard," Revision 2, dated November 2008 (ADAMS Accession No. ML083430462)
| |
| [6] ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum A to RAS-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, dated February 2009
| |
| [7] Quad Cities Nuclear Generating Station Units 1 and 2 PRA Peer Review Report Using ASME/ANS PRA Standard Requirements, BWROG, dated April 2017
| |
| [8] Risk Management Finding Level F&O Independent Assessment Quad Cities Units 1 and 2, 032466-RPT-001, Revision 1, dated May 2021
| |
| [9] Quad Cities PRA Focused-Scope Peer Review, 032466-RPT-02, Revision 0, dated April 2021
| |
| [10] Quad Cities PRA Finding Level Fact and Observation Independent Assessment, 032466-RPT-007, dated June 2021 E2-4
| |
| | |
| License Amendment Request Enclosure 2 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2
| |
| [11] Quad Cities Nuclear Power Station (QCNPS) Unit 1 Fire PRA Peer Review Report Using ASME/ANS PRA Standard Requirements, dated September 2013
| |
| [12] NEI 07-12, "Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines," Revision 1, dated June 2010 (ADAMS Accession No. ML102230070)
| |
| [13] Risk Management Focused-Scope Peer Review Quad Cities Unit 1, 032466-RPT-008, dated June 2021
| |
| [14] NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," Revision 0-A, dated October 12, 2012 (ADAMS Accession No. ML12286A322)
| |
| [15] Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,'
| |
| dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| [16] Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, January 2018 (ADAMS Accession No. ML17317A256)
| |
| E2-5
| |
| | |
| ENCLOSURE 3 License Amendment Request Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Information Supporting Technical Adequacy of PRA Models Without PRA Standards Endorsed by Regulatory Guide 1.200, Revision 2 This enclosure is not applicable to the Quad Cities Nuclear Power Station submittal.
| |
| Constellation is not proposing to use any PRA models in the QCNPS Risk Informed Completion Time Program for which a PRA standard, endorsed by the NRC in RG 1.200, Revision 2 does not exist.
| |
| | |
| ENCLOSURE 4 License Amendment Request Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models 1 Introduction and Scope Topical Report NEI 06-09, Revision 0-A (Reference [1]), as clarified by the U.S. Nuclear Regulatory Commission (NRC) final safety evaluation (Reference [2]), requires that the License Amendment Request (LAR) provide a justification for exclusion of risk sources from the Probabilistic Risk Assessment (PRA) model based on their insignificance to the calculation of configuration risk as well as discuss conservative or bounding analyses applied to the configuration risk calculation. This enclosure addresses this requirement by discussing the overall generic methodology to identify and disposition such risk sources. This enclosure also provides the Quad Cities Nuclear Power Station (QCNPS) specific results of the application of the generic methodology and the disposition of impacts on the QCNPS Risk Informed Completion Time (RICT) Program.
| |
| Section 3 of this enclosure presents the plant-specific analysis of seismic risk to QCNPS.
| |
| Section 4 of this enclosure presents the justification for excluding analysis of high wind risk to QCNPS. Section 5 presents the justification for excluding external flooding risk for QCNPS.
| |
| Section 6 of this enclosure presents the justification for excluding analyses of other external hazards from the QCNPS PRA.
| |
| Topical Report NEI 06-09 does not provide a specific list of hazards to be considered in a RICT Program. However, non-mandatory Appendix 6-A in the ASME/ANS PRA Standard (Reference [3]) provides a guide for identification of most of the possible external events for a plant site. Additionally, NUREG-1855 (Reference [4]) provides a discussion of hazards that should be evaluated to assess uncertainties in plant PRAs and support the risk-informed decision-making process. This information was reviewed for the QCNPS site and augmented with a review of information on the site region and plant design to identify the set of external events to be considered. The information in the Updated Final Safety Analysis Report (UFSAR) regarding the geologic, seismologic, hydrologic, and meteorological characteristics of the site region as well as present and projected industrial activities in the vicinity of the plant were also reviewed for this purpose. No new site-specific and plant-unique external hazards were identified through this review. The list of hazards in Appendix 6-A of the PRA Standard were considered for QCNPS as summarized in Table E4-16.
| |
| The scope of this enclosure is consideration of the hazards in Table E4-16 for QCNPS. As explained in subsequent sections of this enclosure, risk contribution from seismic, high wind, and external flooding events are evaluated quantitatively, and the other listed external hazards are evaluated and screened as having low risk.
| |
| 2 Technical Approach The guidance contained in NEI 06-09 states that all hazards that contribute significantly to incremental risk of a configuration must be quantitatively addressed in the implementation of the RICT Program. The following approach focuses on the risk implications of specific external hazards in the determination of the risk management action time (RMAT) and RICT for the Technical Specification (TS) Limiting Conditions for Operation (LCOs) selected to be part of the RICT Program.
| |
| E4-2
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Consistent with NUREG-1855 (Reference [4]), external hazards may be addressed by:
| |
| : 1) Screening the hazard based on a low frequency of occurrence,
| |
| : 2) Bounding the potential impact and including it in the decision-making, or
| |
| : 3) Developing a PRA model to be used in the RMAT/RICT calculation.
| |
| The overall process for addressing external hazards considers two aspects of the external hazard contribution to risk.
| |
| The first is the contribution from the occurrence of beyond design basis conditions, e.g.,
| |
| winds greater than design, seismic events greater than the design-basis earthquake (DBE), etc. These beyond design basis conditions challenge the capability of the structures, systems, and components (SSCs) to maintain functionality and support safe shutdown of the plant.
| |
| The second aspect addressed is the challenges caused by external conditions that are within the design basis, but still require some plant response to assure safe shutdown, e.g., high winds or seismic events causing loss of offsite power, etc. While the plant design basis assures that the safety related equipment necessary to respond to these challenges are protected, the occurrence of these conditions nevertheless causes a demand on these systems that present a risk.
| |
| Hazard Screening The first step in the evaluation of an external hazard is screening based on an estimation of a bounding core damage frequency (CDF) for beyond design basis hazard conditions. An example of this type of screening is reliance on the NRCs 1975 Standard Review Plan (SRP)
| |
| (Reference [5]), which is acknowledged in the NRCs Individual Plant Examination of External Events (IPEEE) procedural guidance (Reference [6]) as assuring a bounding CDF of less than 1E-6/yr for each hazard. The bounding CDF estimate is often characterized by the likelihood of the site being exposed to conditions that are beyond the design basis limits and an estimate of the bounding conditional core damage probability (CCDP) for those conditions. If the bounding CDF for the hazard can be shown to be less than 1E-6/yr, then beyond design basis challenges from that hazard can be screened out and do not need to be addressed quantitatively in the RICT Program.
| |
| The basis for this is as follows:
| |
| The overall calculation of the RICT is limited to an incremental core damage probability (ICDP) of 1E-5.
| |
| The maximum time interval allowed for this RICT is 30 days.
| |
| If the maximum CDF contribution from a hazard is <1E-6/yr, then the maximum ICDP from the hazard is <1E-7 (1E-6/yr
| |
| * 30 days/365 days/yr).
| |
| E4-3
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Thus, the bounding ICDP contribution from the hazard is shown to be less than 1% of the permissible ICDP in the bounding time for the condition. Such a minimal contribution is not significant to the decision in computing a RICT.
| |
| The QCNPS IPEEE hazard screening analysis (Reference [7]) has been updated to reflect current QCNPS site conditions. The results are discussed in Section 6 and show that all the events listed in Table E4-16 can be screened except seismic events for QCNPS.
| |
| Hazard Analysis - CDF There are two options in cases where the bounding CDF for the external hazard cannot be shown to be less than 1E-6/yr. The first option is to develop a PRA model that explicitly models the challenges created by the hazard and the role of the SSCs included in the RICT Program in mitigating those challenges. The second option for addressing an external hazard is to compute a bounding CDF contribution for the hazard.
| |
| Evaluate Bounding LERF Contribution The RICT Program requires addressing both core damage and large early release risk. When a comprehensive PRA does not exist, the large early release frequency (LERF) considerations can be estimated based on the relevant parts of the internal events LERF analysis. This can be done by considering the nature of the challenges induced by the hazard and relating those to the challenges considered in the internal events PRA. This can be done in a realistic manner or a conservative manner. The goal is to provide a representative or bounding conditional large early release probability (CLERP) that aligns with the bounding CDF evaluation. The incremental large early release frequency (ILERF) is then computed as follows:
| |
| ILERFHazard = ICDFHazard
| |
| * CLERPHazard The approach used for seismic LERF is described in Section 3.
| |
| Risks from Hazard Challenges Given the selection of an estimated bounding CDF/LERF, the approach considered must assure that the RICT Program calculations reflect the change in CDF/LERF caused by the out of service equipment. For QCNPS, as discussed later in this enclosure, the only beyond design basis hazard that could not be screened out are the seismic and extreme wind (tornado missiles only) hazards, and the approach used considers that the change in risk with equipment out of service will not be higher than the estimated seismic CDF.
| |
| The above steps address the direct risks from damage to the facility from external hazards.
| |
| While the direct CDF contribution from beyond design basis hazard conditions can be shown to be non-significant using these steps without a full PRA, there are risks that may be addressed.
| |
| These risks are related to the fact that some external hazards can cause a plant challenge even for hazard severities that are less than the design basis limit. For example, high winds, tornadoes, and seismic events below the design basis levels can cause extended loss of offsite E4-4
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models power conditions. Additionally, depending on the site, external floods can challenge the availability of normal plant heat removal mechanisms.
| |
| The approach taken in this step is to identify the plant challenges caused by the occurrence of the hazard within the design basis and evaluate whether the risks associated with these events are either already considered in the existing PRA model or they are not significant to risk.
| |
| 3 SEISMIC RISK CONTRIBUTION ANALYSIS 3.1 Introduction A seismic PRA (SPRA) was not developed for the QCNPS Individual Plant Examination for External Events (IPEEE - Reference [7]), an alternative approach is taken to provide an estimate of seismic core damage frequency (SCDF) based on the current QCNPS seismic hazard curve, Reference [8], and using a plant level seismic fragility based on the QCNPS IPEEE seismic margins analysis (Reference [7]) and the NRC GI-199 risk assessment report (Reference [9]).
| |
| The calculation of SLERF is performed by estimating an average seismic conditional Large Early release probability (SCLERP), based on the spectrum of SCDF accident sequence types, and multiplying SCDF by the average SCLERP estimate.
| |
| 3.2 Assumptions and Ground Rules
| |
| : 1. Hazard Curve: The QCNPS seismic hazard is defined by the seismic hazard curve (SHC) provided to the NRC in Reference [10] and developed per the seismic hazard analysis documented in Reference [8].
| |
| : 2. PGA Metric: The ground motion metric used to define the seismic hazard in this analysis is peak ground acceleration (PGA). PGA is a common ground motion metric used in seismic risk assessment analyses for nuclear power plants (Reference [11]). Per NRC preference as evidenced in RAIs to industry RICT LARs, this analysis also assesses other hazard metrics (1.0 Hz, 2.5 Hz, 5 Hz, and 10 Hz) and concludes the PGA hazard is reasonable for use in RICT seismic risk estimation.
| |
| : 3. Plant Level Seismic Fragility: The assumed limiting plant seismic capacity used in the QCNPS seismic penalty calculation has a high confidence of low probability of failure (HCLPF) value of 0.24g PGA. The HCLPF capacity is defined as the earthquake level in which there is 95% confidence of less than 5% seismic-induced failure probability. The bases for this value are:
| |
| - QCNPS IPEEE Seismic Margins Assessment (Reference [7])
| |
| - NRC evaluation report (SER) of the QCNPS IPEEE (Reference [12]) which acknowledges that the QCNPS plant level HCLPF capacity is at least 0.24g PGA following IPEEE outlier resolutions and plant improvements (e.g., enhanced anchorage/support, replacement of specific relays).
| |
| E4-5
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models NOTE: the NRC in Reference [9] [GI-199, Table B.2], cites a QCNPS plant level seismic HCLPF of 0.09g PGA. This value is the lowest HCLPF calculated in the QCNPS IPEEE submittal but is not the seismic plant level HCLPF following IPEEE outlier resolutions.
| |
| - QCNPS NTTF 2.3 seismic walkdown submittal (Reference [13]); Section 7 and Appendix G of Reference [13] document that all QCNPS IPEEE outlier resolutions and improvements have been completed.
| |
| The IPEEE assessed QCNPS SSCs associated with the SMA (seismic margins analysis) modeled accident sequences to a review level earthquake (RLE) value of 0.30g PGA in accordance with NRC guidance in NUREG-1407 (Reference [6]). The EPRI seismic margin assessment methodology (Reference [14]) was used for the QCNPS IPEEE seismic analysis.
| |
| An RLE is an earthquake larger than the plant safe shutdown earthquake (SSE) that is chosen for use in SMA for SSC seismic capacity screening purposes to demonstrate sufficient actual margin over the SSE. Typically, the RLE is defined in terms of a ground motion spectrum. The term seismic margins earthquake (SME) is also used.
| |
| The QCNPS IPEEE (Reference [7]) and the NRC SER (Reference [12]) were reviewed for insights to determine the limiting plant HCLPF. For QCNPS, the RLE shape assigned by the NRC is the median NUREG/CR-0098 (Reference [15]) spectrum anchored at 0.30g PGA. The IPEEE included a review of the integrity of the containment itself, containment isolation systems such as valves, mechanical and electrical penetrations, bypass systems, and plant-unique systems such as active seals. Failure of the containment structure and suppression chamber were screened out based on EPRI SMA guidance. The IPEEE did not identify any containment vulnerabilities with respect to seismic events.
| |
| The uncertainty parameter for the plant level seismic capacity in this analysis is represented by a composite beta factor (c) of 0.4. This is a commonly accepted approximation and is consistent with the value used in GI-199, Table C.1, Bases for Establishing Plant-Level Fragility Curves Parameters from IPEEE Information (Reference [9]). Refer to Section 4 of Reference [11], as necessary, for further discussion of fragility uncertainty parameters.
| |
| : 4. Convolution to Determine SCDF: The estimation of SCDF in this calculation is performed by a mathematical convolution of various QCNPS seismic hazard curves (i.e., PGA and other spectral acceleration curves from Reference [8]) and the SMA-based plant level seismic fragility. This convolution estimation approach is a common analysis in approximating an SCDF for use in risk-informed decision making (e.g., it is commonly used in RICT seismic penalty calculations; the NRC used this approach in the GI-199 risk assessment in Reference [9]) in absence of a current full-scope SPRA.
| |
| : 5. SLERF: The QCNPS SLERF for this risk evaluation is obtained by multiplying the calculated SCDF by an average seismic conditional large early release probability (SCLERP). The average SCLERP is estimated using information from both the QCNPS FPIE PRA (Reference [16]) and fragility information from industry SPRAs.
| |
| E4-6
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models 3.3 Calculations The general approach to estimation of the SCDF is to use the plant level seismic fragility and convolve the corresponding failure probabilities as a function of seismic hazard level with the seismic hazard curve frequencies of occurrence. This is a commonly used approach to estimate SCDF when a seismic PRA is not available. This approach is the same as that used in the Vogtle pilot TSTF-505 license amendment request submittal (Reference [17]) and previous Exelon /
| |
| Constellation TSTF-505 submittals, e.g., for Calvert Cliffs (Reference [18]).
| |
| The key elements of the SCDF convolution calculation (i.e., seismic hazard curve and associated hazard intervals used in the convolution calculation; plant level seismic-induced failure probabilities based on the plant level seismic fragility; and the resulting SCDF from the convolution calculation) and the SLERF calculations are discussed in the next sub-Sections.
| |
| 3.4 Seismic Hazard and Intervals The seismic hazard in units of g (PGA, peak ground acceleration) is shown in Table E4-1 (from Reference [10] with the hazard curves taken from Reference [8].) The mean fractile occurrence frequencies of Table E4-1 are used in the calculations here; use of mean values is a typical and expected PRA practice.
| |
| The convolution calculation (summarized later in Table E4-2) of the seismic hazard curve with the QCNPS PGA-based plant level seismic fragility curve is performed by dividing the hazard curve into seismic magnitude range intervals. In the case of the seismic hazard curve in Table E4-1, nineteen seismic hazard intervals are explicitly used in this convolution calculation and are defined by the magnitude data points (0.0005 to 0.001g; 0.001g to 0.005; and ultimately to
| |
| >10g).
| |
| To facilitate calculation of the QCNPS plant fragility probability at each seismic hazard interval, a representative g-level is calculated for each interval. The representative g-level for the seismic hazard intervals is calculated using a geometric mean approach (i.e., the square root of the product of the g-level values at the beginning and end of a given interval). For the last open-ended seismic interval greater than 10g, the representative g-level is estimated as 11g.
| |
| However, the precision of the representative magnitude used for the final open-ended seismic interval in the SCDF convolution is immaterial given that the calculated conditional failure probability of the final hazard interval (as well as the preceding three hazard intervals) is 1.0 and the contribution from this final interval has a negligible contribution to the overall SCDF estimate.
| |
| The seismic hazard interval annual initiating event frequency is calculated (except for the final interval) by subtracting the mean exceedance frequency associated with the g-interval (high) end point from the mean exceedance frequency associated with the g-interval beginning point.
| |
| The frequency of the last seismic hazard interval is the exceedance frequency at the beginning point of that interval. This is common practice in industry SPRAs (Reference [11]).
| |
| E4-7
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-1: EPRI 2014 Seismic Hazard Data for QCNPS1 AMPS(g) Mean 0.05 0.16 0.50 0.84 0.95 0.0005 5.86E-02 2.32E-02 4.07E-02 5.83E-02 7.77E-02 8.85E-02 0.001 3.88E-02 1.38E-02 2.46E-02 3.73E-02 5.42E-02 6.64E-02 0.005 9.13E-03 2.92E-03 4.98E-03 8.12E-03 1.25E-02 1.98E-02 0.01 4.26E-03 1.20E-03 1.95E-03 3.57E-03 6.09E-03 1.05E-02 0.015 2.46E-03 6.54E-04 9.79E-04 1.87E-03 3.57E-03 7.03E-03 0.03 7.92E-04 1.69E-04 2.64E-04 5.05E-04 1.05E-03 2.84E-03 0.05 3.21E-04 5.58E-05 9.51E-05 1.92E-04 4.43E-04 1.20E-03 0.075 1.57E-04 2.46E-05 4.50E-05 9.37E-05 2.29E-04 5.58E-04 0.1 9.50E-05 1.44E-05 2.72E-05 5.91E-05 1.42E-04 3.19E-04 0.15 4.71E-05 6.64E-06 1.32E-05 3.01E-05 7.23E-05 1.46E-04 0.3 1.33E-05 1.44E-06 3.28E-06 8.72E-06 2.10E-05 3.90E-05 0.5 4.65E-06 3.28E-07 8.98E-07 2.92E-06 7.77E-06 1.42E-05 0.75 1.84E-06 7.66E-08 2.60E-07 1.04E-06 3.19E-06 6.09E-06 1 8.96E-07 2.25E-08 9.24E-08 4.56E-07 1.57E-06 3.19E-06 1.5 2.96E-07 3.14E-09 1.74E-08 1.21E-07 5.27E-07 1.15E-06 3 3.30E-08 1.32E-10 6.17E-10 7.55E-09 5.27E-08 1.44E-07 5 4.87E-09 1.01E-10 1.11E-10 6.64E-10 6.64E-09 2.25E-08 7.5 8.65E-10 9.11E-11 1.01E-10 1.42E-10 1.04E-09 4.07E-09 10 2.25E-10 8.12E-11 9.11E-11 1.02E-10 2.88E-10 1.11E-09 3.5 Seismic Failure Probabilities The seismic failure probability of the QCNPS limiting plant fragility for each seismic hazard interval is calculated using the following fragility equations (this is for the mean confidence level). These are the typical lognormal fragility equations used in most hazard PRAs (Reference [11]).
| |
| Fragility (i.e., failure probability) = [ln(A/Am)/c], where:
| |
| is the standard lognormal distribution function A is the g level in question, Am is the median seismic capacity, and the uncertainty parameters (betas) are related as follows:
| |
| c = (u2 + r2)0.5 HCLPF and Am are related as follows:
| |
| Am = HCLPF / (exp -2.33c)
| |
| As discussed previously, the HCPLF point of the QCNPS IPEEE plant level seismic fragility curve is 0.24g PGA. The uncertainty variable c for the QCNPS plant level fragility is set to a 1
| |
| From Reference [8] Table A-1a. Mean and Fractile Seismic Hazard Curves for PGA (100 Hz) at QCNPS E4-8
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models value of 0.4; this uncertainty variable value is consistent with that used by the NRC for the QCNPS plant in Reference [9] as well as it typical for use as a representative composite uncertainty (refer to Section 6.4 of Reference [19]).
| |
| 3.6 Seismic Core Damage Frequency The SCDF for each hazard interval is computed as the product of the hazard interval initiating event frequency (/yr.) and the plant level seismic fragility failure probability for that same hazard interval. The results per hazard interval are then straight summed to produce the overall total SCDF across the entire hazard curve. The SCDF convolution calculation is summarized in Table E4-2 and shows the total estimated SCDF is 4.3E-6/yr. Table E4-2 provides the following information:
| |
| Seismic hazard intervals and their associated initiating event frequencies (Mean) and representative magnitudes.
| |
| Plant level HCLPF fragility failure probabilities (Mean) per hazard interval.
| |
| Convolved SCDF per interval and total SCDF.
| |
| To evaluate the effect of using the different available hazard curves from the QCNPS PSHA, the seismic penalty calculation has been re-performed with different hazard curves in the 1 Hz to 10 Hz range (i.e., the range of spectral values that would be used in an SPRA as an alternative to PGA). The plant-level fragility median (Am) value is adjusted per the QCNPS 2014 ground motion response spectra shape (Table 2.4-1 of Reference [8]) for each convolution sensitivity case (keeping c=0.4 for each case).
| |
| The convolved SCDF result from each hazard curve is summarized below:
| |
| PGA 10 Hz 5 Hz 2.5 Hz 1 Hz Plant-Level Fragility (Am): 0.61g 1.30g 0.90g 0.42g 0.29g Convolved SCDF: 4.3E-6 4.3E-6 4.0E-6 3.7E-6 3.2E-6 As can be seen, the resulting convolved SCDF using PGA is the highest of the five Hz cases (note: the 10 Hz-based SCDF is slightly lower in the second decimal place than the PGA-based SCDF); this is not uncommon that PGA produces the highest SCDF. The PGA-based SCDF (and SLERF) are used for the QC RICT seismic penalty calculation.
| |
| 3.7 Seismic Large Early Release Frequency For use of a seismic penalty estimate in RICT calculations, it can be unacceptably conservative to simply assume that Delta SCDF = Delta SLERF. As such, a less conservative approach is pursued here for the QCNPS SLERF estimation. The QCNPS SLERF is determined in this calculation by multiplying the estimated SCDF shown in Table E4-2 (4.3E-6/yr.) by an average seismic conditional large early release probability (SCLERP). An estimate of the average SCLERP is calculated using 1) an estimation of the breakdown of SCDF by accident sequence type and 2) PRA accident sequence progression information from the quantification results of E4-9
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models the current QCNPS FPIE PRA model of record (Reference [16]) adjusted to reflect the influence of seismic-induced failures.
| |
| This SLERF methodology is discussed below according to the following topics:
| |
| Spectrum of seismic-induced core damage accident sequence types CLERP as a function of seismic core damage accident sequence type Application of SLERF in RICT Calculations Spectrum of Seismic-Induced Core Damage Accident Sequence Types The estimation of an average SCLERP requires as an input the assessment of the contribution of different accident sequence types to seismic core damage frequency (SCDF). The contribution of various accident sequence types (or accident classes) to core damage frequency at a given plant is not necessarily the same between FPIE PRA and other hazard (e.g., seismic)
| |
| PRAs. Given that QCNPS does not have a detailed plant-specific seismic PRA to assist in estimating the spectrum of SPRA accident sequence types and SCLERP, the range of fragility information from other recently submitted SPRAs is reviewed to assist in this calculation.
| |
| Based on past and current SPRAs (e.g., References [20], [21], [22], [23], [24], [25], [26], [27],
| |
| [28], [29], [30], [31], [32], [33], [34]), the spectrum of SCDF sequence types are as follows (they cover the typical BWR PRA core damage accident classes: I, II, III, IV and V, e.g., refer to Table B-1 of Reference [35]):
| |
| Seismic-LOOP with early loss of RPV injection: These are seismic-induced loss of offsite power scenarios with RPV coolant injection failure at t=0.
| |
| Seismic-LOOP with delayed loss of RPV injection: These are seismic-induced loss of offsite power scenarios with RPV coolant injection initially running but subsequently failed (e.g., due to loss of battery charging and subsequent depletion of DC supply).
| |
| As discussed later, loss of decay heat removal scenarios are included in this category in this analysis.
| |
| Seismic-LOOP with Seismic-LOCA: These are scenarios with a seismic-induced LOOP and seismic-induced LOCA (small, medium or large). This category would include loss of injection scenarios and loss of containment heat removal scenarios.
| |
| Seismic-ATWS Unmitigated: These are seismic-induced failure to scram scenarios with failure of reactivity control (e.g., failure of standby liquid control). These accidents proceed with high reactor power discharge into primary containment resulting in dynamic loading and failure of the primary containment structure.
| |
| Adequate core cooling is subsequently failed due to primary containment failure and the subsequent effect on RPV injection equipment. [Note: Given the very low probability of random (i.e., non-seismic induced) failure to scram, seismic-induced accident sequences with random failure to scram are encompassed by this accident class category.]
| |
| E4-10
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-2: Convolution Calculation Summary of QCNPS Seismic CDF Hazard Interval Hazard Mean Hazard Interval Convolved Peak Ground Representative Interval Exceedance Occurrence Frequency Acceleration (g) Magnitude Fragility Frequency (/yr) Frequency (/yr) (/yr)
| |
| (Geo. Mean, g PGA) (Mean) 0.0005 5.86E-02 0.0007 2.33E-64 1.98E-02 4.61E-66 0.001 3.88E-02 0.0022 5.90E-45 2.97E-02 1.75E-46 0.005 9.13E-03 0.0071 3.94E-29 4.87E-03 1.92E-31 0.01 4.26E-03 0.0122 7.70E-23 1.80E-03 1.39E-25 0.015 2.46E-03 0.0212 2.33E-17 1.67E-03 3.88E-20 0.03 7.92E-04 0.0387 2.79E-12 4.71E-04 1.31E-15 0.05 3.21E-04 0.0612 4.60E-09 1.64E-04 7.55E-13 0.075 1.57E-04 0.0866 5.35E-07 6.20E-05 3.32E-11 0.1 9.50E-05 0.1225 3.01E-05 4.79E-05 1.44E-09 0.15 4.71E-05 0.2121 4.16E-03 3.38E-05 1.41E-07 0.3 1.33E-05 0.3873 1.28E-01 8.65E-06 1.11E-06 0.5 4.65E-06 0.6124 5.05E-01 2.81E-06 1.42E-06 0.75 1.84E-06 0.8660 8.10E-01 9.44E-07 7.65E-07 1 8.96E-07 1.2247 9.59E-01 6.00E-07 5.76E-07 1.5 2.96E-07 2.1213 9.99E-01 2.63E-07 2.63E-07 3 3.30E-08 3.8730 1.00E+00 2.81E-08 2.81E-08 5 4.87E-09 6.1237 1.00E+00 4.01E-09 4.00E-09 7.5 8.65E-10 8.6603 1.00E+00 6.40E-10 6.40E-10 10 2.25E-10 11.0000 1.00E+00 2.25E-10 2.25E-10 Total Convolved SCDF Across PGA Hazard Curve (1/yr.): 4.3E-6 Direct to Core Damage: These are scenarios with significant seismic-induced failures that are modeled directly as core damage. Such scenarios include LOCA outside containment scenarios and key structural failures, e.g.: RPV support failure, reactor building and control building (Service Building at QCNPS) structural failure. This category of accidents can be further subdivided as:
| |
| - DCDLRF: Seismic-induced failures that are often modeled in SPRAs directly as SCDF and SLERF (e.g., containment structural failures, containment bypass scenarios).
| |
| - DCD: Seismic-induced failures that are often modeled in SPRAs directly as SCDF but not necessarily direct to SLERF (e.g., failure of a control building would effectively create a loss of injection at t=0 scenario but such a scenario would not lead directly to LERF).
| |
| E4-11
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Seismic-Transients (no LOOP): These seismic-initiated accident sequences have offsite power available and thus potential use of balance of plant (BOP) equipment, e.g., Feedwater This category can include loss of coolant injection scenarios and loss of containment heat removal scenarios. Given the low seismic-capacity of offsite power equipment in comparison to the much higher seismic capacities of the NSSS system, the scram system, and other safety equipment, the contribution from seismic-transient scenarios would be a low contributor to SCDF (e.g., a couple percentage points).
| |
| The above accident sequence categories cover the key critical safety functions (reactivity control, core cooling, RPV and containment integrity) and are sufficient to describe the spectrum of SCDF accident sequences. Determining the estimated fractional contribution to SCDF from each of the accident sequence types is approached as follows:
| |
| Similar to a seismic initiating event tree structure, consider the more severe sequence types first and the remainder proceed to lesser severe accident states.
| |
| Information from industry SPRAs is used to inform the selection of key SSC fragilities.
| |
| Figure E4-1 shows in graphical format the approach to estimating the fraction contribution to SCDF by accident sequence type. Figure E4-1 is a seismic-initiating event tree that begins with severe seismic-induced failures that would typically be modeled as directly to SCDF and SLERF (i.e., the down branch of the No Direct LERF, S-DCDLRF, node) or directly to SCDF but not directly to SLERF (i.e., the down branch of the "No Direct CD", S-DCD, node). The next node, S-LOOP, models seismic-induced loss of offsite power. The next node, S-ATWS), models seismic-induced failure to scram and the final two nodes (S-LOCA and S-HPI) model seismic-induced LOCA and seismic-induced failure of high-pressure injection, respectively.
| |
| A convolution calculation of the Table E4-1 seismic hazard curve and the fragility Am value shown at the branch points is performed for each of the accident sequence type categories shown in the SIET of Figure E4-1:
| |
| S-DCDLRF Node: Down at this node produces an accident sequence type that is typically and reasonably modeled in an SPRA as directly to core damage and directly to LERF (thus, CLERP=1.0). The SSC fragility for this node are the QCNPS primary containment structure or other significant failures that would result in primary containment bypass (e.g., RPV supports for a BWR). A median capacity of 1.5g PGA is selected to model this node. This capacity is reasonably judged to be on the low end of the capacity range given the range of fragilities observed based on review of recent and past SPRAs. The uncertainty parameter for seismic capacity is represented by a composite beta factor (c) of 0.4; as discussed in Section 3.5 which is a commonly accepted approximation.
| |
| S-DCD Node: Down at this node produces an accident sequence type that is typically and reasonably modeled in an SPRA as directly to core damage but not typically directly to LERF. The SSC fragility for this node are structures such as a diesel generator E4-12
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models building or control building (in the case of QCNPS the Services Building), TB hardened areas for safety switchgear and batteries). Such failures in conjunction with assuming no recovery potential prior to the onset of core damage would result directly in core damage but the conditional probability of producing LERF would be less than 1.0. A median capacity of 1g PGA is selected to model this node. This capacity is reasonably judged to be on the low end of the capacity range given the range of fragilities based on a review of recent and past SPRAs. The uncertainty parameter for seismic capacity is represented by a composite beta factor (c) of 0.4; as discussed in Section 3.5. This is a commonly accepted approximation. The convolution calculation for accident sequence type #6 (S-DCD) in Figure E4-1 includes the success probability for the up-branch of the previous node.
| |
| S-LOOP Node: If the S-DCDLRF and S-DCD failures did not occur this node then considers seismic-induced loss of offsite power. The seismic capacity of the offsite power function is assigned an Am=0.3g PGA capacity, with uncertainties r=0.3 and u=0.45 (based on Table A-0-4 of Reference [36]).
| |
| S-ATWS Node: This node models the probability of a seismic-induced failure to scram due to seismic-induced failure of reactor internals causing mechanical failure of control rods to insert. A median capacity of 1.25g PGA is selected to model this node. This capacity is reasonably judged to be on the low end of the capacity range given the range of fragilities observed based on a review of recent and past SPRAs. The uncertainty parameter for seismic capacity is represented by a composite beta factor (c) of 0.4; as discussed in Section 3.5 which is a commonly accepted approximation. The convolution calculation for accident sequence type #5 (S-ATWS) in Figure E4-1 includes the success probability for the up-branch of the previous nodes.
| |
| S-LOCA Node: This node models the probability of a seismic-induced LOCA. A median capacity of 1.00g PGA with uncertainties r=0.3 and u=0.4 (based on Small LOCA in Table H-2 of Reference [11]) is selected to model this node. The convolution calculation for accident sequence type #4 (S-LOOP-LOCA) in Figure E4-1 includes the success probability for the up-branch of the previous nodes.
| |
| S-HPI Node: This node models the probability of seismic-induced failure of HPCI and RCIC at t=0. A HCLPF capacity of 0.3g PGA with composite beta factor (c) of 0.4 (which produces a median Am=0.76g PGA) is selected to model this node. This capacity is assigned based on the RLE used in the QC IPEEE SMA, i.e., review of the results of the QC IPEEE SMA (Table 3-3 of Reference [7]) show that the SSCs necessary for HPCI or RCIC initial operation (until battery depletion) all were assessed to meet the 0.3g PGA RLE. As discussed in Section 3.5, a composite beta factor (c) of 0.4 is a commonly accepted approximation. The convolution calculation for accident sequence type #3, (S-LOOP-IBE) in Figure E4-1 includes the success probability for the up-branch of the previous nodes.
| |
| Sequence Type #2 (S-LOOP-IBL): If the S-DCDLRF, S-DCD, S-ATWS, S-LOCA and S-HPI failures did not occur then the potential remaining sequence category is a E4-13
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models seismic-induced LOOP state with high pressure injection not directly failed. The fractional SCDF risk contribution for sequence type #2 (S-LOOP-IBL) is calculated as 1.0 minus the sum of the fractional contributions from the other sequence types. The convolution calculation for accident sequence type #2 (S-LOOP-IBL) in Figure E4-1 includes the success probability for the up-branch of the previous nodes.
| |
| The above approach to estimating the fractional contribution of accidents sequence types to SCDF is reasonable and conservative for this purpose of seismic penalty RICT calculation.
| |
| Two key conservatisms (with respect to the impact on the resulting average SCLERP estimate) are as follows:
| |
| The lower range of seismic capacity is assigned to the nodal branch points in Figure E4-1 to force higher fractional contributions to those accident sequence types with higher CLERPs.
| |
| Direct summation and subtraction, as opposed to Boolean algebra, is used to determine the accident sequence contributions to SCDF. If a Boolean summation approach were used the risk contribution from S-DCD and S-DCDLRF in a full SPRA would be much less than the combined contribution of 40% shown in Figure E4-1 and used in this risk calculation. This can be illustrated by an example from the Dresden SPRA. The Dresden SPRA fragility screening level of 1g HCLPF is shown in Appendix C of Reference [27] to be 2.1% of SCDF when using straight math estimation of the contribution to SCDF; however, when the screening fragility group is added directly into the Dresden SPRA single-top fault tree logic model and the importance calculated using Boolean algebra the fractional contribution from the 1g HCLPF screening fragility group is shown to be significantly less at <0.5%.
| |
| E4-14
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models FRACTION of NO DIRECT Successful HPCI, RCIC SCOF SEISMIC EVENT NO DIRECT CD Offslte AC Intact No LOCA # ACCIDENT 1YPE LERF SCRAM Functional (Nominal Fraction)
| |
| S-IE S-DCOLRF S-OCO S-LOOP S-AlWS S-LOCA S-HPI 1 S-TRANS non-significant 2 S-LOOP-IBL 0.20 Am (g) = 076g 3 S-LOOP- IBE 0.25 Am (g) =0.3 Am (g)= 1.0 4 S-LOOP-LOCA 0.10 Am (g)= 1.25 5 S-A1WS 0.05 Am (g)= 1.00 6 S-OCD 0.25 Am {g): 1.50 7 S-DCDLRF 0.15 Figure E4-1: Contribution of SCDF by Accident Sequence Type E4-15
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models CLERP as a Function of SCDF Accident Sequence Type The next step in the estimation of an average seismic CLERP is to estimate the CLERP for each SCDF accident sequence type. A given accident sequence type may not result in a core damage event until well after the PRA "early" release time frame (defined in the QCNPS FPIE PRA as < 5 hours from the time of the cue for a General Emergency declaration; per Table 5.4.3-1 of Reference [37]. Conversely, some accident sequence types would, by PRA convention, be modeled directly as a LERF, such as a station blackout scenario with no initial RPV injection and failure to manually isolate containment isolation valves that are initially open and do not automatically isolate.
| |
| Seismic CLERP as a function of SCDF accident sequence type is summarized in Table E4-3 and discussed below (ordered by % contribution, low to high).
| |
| Seismic-induced LOOP with early loss of coolant injection (S-LOOP-IBE) is assigned an SCLERP of 3.6E-01. This SCLERP estimate is based on results from the current QCNPS at-power internal events PRA (Reference [16]) for loss of offsite power scenarios with no RPV coolant injection at t=0 and no recovery of AC or coolant injection. This CLERP result is applicable to a seismic-induced accident sequence with loss of all injection at t=0 and no recovery given that the probability of the LERF release is a function of the accident progression phenomena, containment failure location (e.g., drywell shell melt-through) and the potential for pre-existing primary containment leakage condition (basic events 1CNHU-PREINITH-- and 1CZPH-LIQ-N2-F--). In addition, refer to later discussion on the negligible likelihood of seismic-induced failure of containment or containment isolation.
| |
| Seismic-induced failures leading directly to core damage but not directly to LERF (S-DCD) are damage states that will eventually lead directly to core damage (whether in early release time frame or later time frames) but the primary containment not seismically failed. The 0.53 SCLERP assigned to the S-DCD accident type is based on assuming 1/3 contribution each for the following three general scenarios: 1) S-LOOP-IBE SCLERP condition of HPCI and RCIC directly failed at t=0 (0.36 SCLERP); 2) EDGs directly failed at t=0 but HPCI and RCIC not directly failed (0.25 SCLERP) and 3) EDGs and HPCI and RCIC directly failed at t=0 (0.98 SCLERP).
| |
| Seismic-induced failures leading directly to core damage and LERF are, by definition, assigned an SCLERP of 1.0. The scenarios include seismic-induced failures such as containment structural failures, RPV support failures, ISLOCA and containment isolation failures.
| |
| Unmitigated ATWS scenarios are assigned an SCLERP of 5.1E-1 based on the results of the QC118A Level 2 PRA results. The QCNPS FPIE-based CLERP for Class IV core damage accidents is 0.51. Based on review of the Class IV LERF cutsets from the QC118A FPIE, the ATWS CLERP is overwhelmingly dominated
| |
| (>98%) by accident phenomena issues such as dynamic loading failure of the wetwell (e.g., basic event 1CNWW-ATWSSEQF--) and primary containment drywell E4-16
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models shell melt-through (basic event 1SIPHCONTFAILF--). These Level 2 failures are directly applicable to the SPRA as they do not take inappropriate probabilistic credit for functions that may be influenced by seismic-induced impacts. An SCLERP of 0.51 is used in the seismic penalty calculation for seismic ATWS CDF.
| |
| Seismic-induced LOOP with initial coolant injection (S-LOOP-IBL) is assigned an SCLERP of 1.3E-01. This SCLERP estimate is based on results from the current QCNPS at-power internal events PRA (Reference [16]) for LOOP with no offsite AC recovery and the run times for emergency AC equipment increased to 24 hrs. in the QC FPIE Level 2 PRA. Per the accident category definition, HPCI and RCIC are not directly failed in this PRA quantification run. The EDGs are not directly failed in this run but a convolution of 0.3g PGA with hazard curve is added to both the CDF and LERF cutsets for this run to simulate seismic fragility of emergency AC.
| |
| Seismic-induced LOOP with initial RPV injection, successful emergency power but subsequent loss of containment heat removal is also conservatively included in this category. S-LOOP with successful continued injection but no containment heat removed would have an SCLERP of 0.05 based on probability of delaying declaration of a General Emergency. Declaration of a General Emergency would be in accordance with QCNPS Emergency Action Levels and if declared per procedure in the appropriate time frame, such scenarios would not result in an "early" release (and thus non-LERF). However, the QCNPS Level 2 PRA includes a 5% probability that the General Emergency declaration is delayed and thus can result in an "early" release for these sequences (refer to Appendix G of the QCNPS Level 2 PRA notebook). Use of a 0.05 SCLERP value would be conservative because it would not account for the containment failure location in reducing release magnitude (i.e., if failure occurs above the suppression pool water line the release would be scrubbed and not a "high" magnitude release). Regardless, a 0.05 SCLERP for S-LOOP with loss of containment heat removal is lower than the S-LOOP-IBL of 0.13 above and this sub-category of accidents is conservatively included in the seismic penalty averaged SCLERP estimate in the S-LOOP-IBL contribution above.
| |
| Seismic-induced LOOP coincident with seismic induced LOCA will result in loss of steam supply for HPCI and RCIC (thus, HPCI and RCIC assumed failed at t=0). The CLERP of 0.36 used for S-LOOP-IBE above reasonably applies here as well. The LOCA condition will reduce the likelihood of certain accident progression energetic phenomena (e.g., in-vessel steam explosions, high pressure melt ejection) but no reduction in SCLERP is applied here for this aspect.
| |
| Seismic-Transients (offsite power available): As summarized previously and in Table E4-3, seismic-induced transients are very low risk contributors to seismic-induced risk and also have a lower CLERP than the other accident sequence types used in the average SCLERP calculation shown in Table E4-3.
| |
| E4-17
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-3: Spectrum of SCDF Accident Sequences and Associated SCLERP L1 SPRA Accident Fraction Of SCLERP(2) Comment Sequence Type SCDF(1)
| |
| S-LOOP-IBE: 0.25 0.36 Based on CLERP results for LOOP with no Seismic-induced loss offsite AC recovery, HPCI and RCIC directly of offsite power and failed at t=0, and the run times for emergency RPV injection AC equipment increased to 24 hrs. in the QC FPIE Level 2 PRA. The EDGs and LPCI/LPCS are not directly failed in this run but a convolution of 0.3g PGA with QCNPS 2014 seismic hazard curve is added to both the CDF and LERF cutsets for this run to simulate seismic fragility of emergency AC.
| |
| S-DCD: Scenarios 0.25 0.53 These direct to core damage (but not direct to direct to core damage SLERF) accident scenarios are seismic-but not direct LERF induced damage states that will eventually (e.g., CB structural lead directly to core damage (whether in early failures or EDG bldg. release time frame or later time frames) but structural failures) primary containment not seismically failed.
| |
| The 0.53 SCLERP assigned to the DCD accident type is based on assuming 1/3 contribution each for the following three general scenarios: 1) S-LOOP-IBE SCLERP condition of HPCI and RCIC directly failed at t=0 (0.36 SCLERP); 2) EDGs directly failed at t=0 but HPCI and RCIC not directly failed (0.25 SCLERP) and 3) EDGs and HPCI and RCIC directly failed at t=0 (0.98 SCLERP).
| |
| S-LOOP-IBL: 0.20 0.13 Based on CLERP results for LOOP with no Seismic-induced loss offsite AC recovery and the run times for of offsite power. emergency AC equipment increased to 24 hrs.
| |
| in the QCNPS FPIE Level 2 PRA. Per the accident category definition, HPCI and RCIC are not directly failed in this PRA quantification run. The EDGs are not directly failed in this run but a convolution of 0.3g PGA with hazard curve is added to both the CDF and LERF cutsets for this run to simulate seismic fragility of emergency AC.
| |
| E4-18
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-3: Spectrum of SCDF Accident Sequences and Associated SCLERP L1 SPRA Accident Fraction Of SCLERP(2) Comment Sequence Type SCDF(1)
| |
| S-DCDLRF: 0.15 1.00 By definition, the accidents scenarios defined Scenarios direct to by seismic-induced failures that would be LERF (e.g., modeled as direct LERF receive an SCLERP containment structural probability of 1.0.
| |
| failures)
| |
| S-LOOP-LOCA: 0.10 0.36 The LOCA condition will result in loss of steam Seismic-induced loss supply for HPCI and RCIC (thus, HPCI and of offsite power and RCIC assumed failed at t=0). The CLERP of seismic-induced 0.36 used for S-LOOP-IBE above reasonably LOCA applies here as well. The LOCA condition will reduce the likelihood of certain accident progression energetic phenomena (e.g., in-vessel steam explosions, high pressure melt ejection) but no reduction in SCLERP is applied here for this aspect.
| |
| S-ATWS: S-ATWS 0.05 0.51 The QCNPS FPIE-based CLERP for Class IV unmitigated core damage accidents is 51%. Based on review of the Class IV LERF cutsets from the QC118A FPIE, the ATWS CLERP is overwhelmingly dominated by accident phenomena issues: 1) primary containment drywell shell melt-through (basic event 1SIPHCONTFAILF--) is approximately 98.6%
| |
| of the CLERP; and 2) other energetic phenomena such as steam explosion (basic event 1CZPH-EXVSLSTF--) or missiles generated pierce drywell (1CZPH-MISDRYWF--) induced failure of primary containment. These Level 2 failures are directly applicable to the SPRA as they do not take inappropriate probabilistic credit for functions that may be influenced by seismic-induced impacts. An SCLERP of 0.51 is used in the seismic penalty calculation for seismic ATWS CDF.
| |
| E4-19
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-3: Spectrum of SCDF Accident Sequences and Associated SCLERP L1 SPRA Accident Fraction Of SCLERP(2) Comment Sequence Type SCDF(1)
| |
| S-LOOP with long Note (3) Note (3) Conservatively treated here by incorporating term loss of this category into the S-LOOP-IBL category.
| |
| containment cooling S-LOOP with successful continued injection but no containment heat removed would have an SCLERP of 0.05 based on probability of delaying declaration of a General Emergency.
| |
| Declaration of a General Emergency would be in accordance with QCNPS Emergency Action Levels and if declared per procedure in the appropriate time frame, such scenarios would not result in an "early" release (and thus non-LERF). However, the QCNPS Level 2 PRA includes a 5% probability that the General Emergency declaration is delayed and thus can result in an "early" release for these sequences (refer to Appendix G of the QCNPS Level 2 PRA notebook). Use of a 0.05 SCLERP value would be conservative because it would not account for the containment failure location in reducing release magnitude (i.e., if failure occurs above the suppression pool water line the release would be scrubbed and not a "high" magnitude release). Regardless, a 0.05 SCLERP for S-LOOP with loss of containment heat removal is lower than the S-LOOP-IBL of 0.13 above and this sub-category of accidents is conservatively included in the seismic penalty averaged SCLERP estimate in the S-LOOP-IBL contribution above.
| |
| S-Transients (no 0.00 n/a S-Transients (no seismic-induced LOOP) are LOOP) reasonably assumed to be non-significant contributors to SCDF and SLERF. This is typical of past SPRAs and due in large part to the comparatively very low seismic capacity of offsite power equipment (primarily ceramic insulators).
| |
| Sequence-Weighted Average 0.46 Sum of (Fraction of SCDF x SCLERP) over all SCLERP: sequence types E4-20
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-3 Notes:
| |
| : 1. Range of key SSC fragilities in industry SPRAs used to inform SCDF breakdown by accident type. Table ordered by sequence % contribution.
| |
| : 2. These are seismically biased CLERP estimates based on QCNPS 2018 FPIE PRA sensitivity quantifications estimates, yielding the "SCLERP" label.
| |
| : 3. These SCDF accident sequence types are variants of the S-LOOP-IBL category above and are addressed in this average SCLERP estimate by conservatively incorporating them into the S-LOOP-IBL category, as summarized in the Comment column.
| |
| The sequence weighted average SCLERP over the SCDF accident sequence contributions and assigned SCLERPs is estimated as 0.46.
| |
| In addition to the average SCLERP estimation discussed previously, the following discussions regarding random and seismic-induced failure of containment isolation failure are provided to support the reasonableness of the average SCLERP estimation (e.g., there are no normally open AC-powered MOV PCIVs that would lead directly to an unscrubbed release and a LERF end state):
| |
| Containment Isolation Random Failure: Random failure of primary containment isolation is already included in the average SCLERP estimation discussed previously. One of the random (non-seismic induced) containment isolation failure contributors is basic event 1CNHU-PREINITH--, PRE EXISTING CONTAINMENT FAILURE.
| |
| Containment Isolation Fragility: Seismic-induced failure of containment isolation is very low likelihood and encompassed by the SCLERPs used in Table E4-3. The containment isolation valves of interest to the LERF risk metric are primarily air-operated valves (AOVs) and motor-operated valves (MOVs), most normally closed at-power, that fail-safe closed on loss of pneumatic or electric power (e.g.,
| |
| seismic-induced LOOP). Successful primary containment isolation in preventing a LERF release for seismic-induced accidents is not dependent upon pneumatic supply, electric power, or containment isolation signals (i.e., ~99% of SCDF involves seismic-induced LOOP and the PCIVs fail-safe closed under such conditions).
| |
| The PCIVs have high seismic capacities such that seismic loading will have a negligible likelihood of failing the PCIVs in the open position. The AOV PCIVs fail-safe closed via internal spring force inside the AOV operator. Once closed, these valves do not need to open again during or after the seismic event.
| |
| Therefore, they do not meet the definition of an "active" valve per the air operated valve equipment class (per the EPRI SQUG Generic Implementation Procedure, GIP, and EPRI NP-7149 Seismic Adequacy of Equipment Classes).
| |
| The spring will successfully cause the PCIVs to shut at accelerations much greater than those associated with the functional failure capacity used to E4-21
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models determine the fragility of active valves. As such, these PCIVs are essentially inactive valves, which are inherently rugged as there is not a credible seismic failure mechanism that would prevent the valves from failing shut as desired. In addition, both in-series AOV PCIVs in a penetration line would have to seismically fail to fail-safe closed to result in an open release pathway.
| |
| Some primary containment penetrations use motor operated valves (MOV) for containment isolation which would require electric power for closure and for an isolation signal. However, such PCIV MOVs are not significant to LERF for one or more of the following reasons:
| |
| MOV in closed position during at-power operation and at the time of the seismic event (e.g., main steam line drains)
| |
| Very small line (e.g., 1-inch diameter instrument gas line)
| |
| AOV or check valve PCIV in-series with the MOV Penetration is a closed-loop system or otherwise scrubbed that would not represent a LERF (i.e., High magnitude release).
| |
| Summary of Seismic LERF Calculation Based on the information in Table E4-3, an estimate of average CLERP for use in seismic penalty RICT calculations is 0.46. Therefore, a seismic penalty of SLERF is calculated as:
| |
| SLERF = 4.31E-6/yr. (SCDF from Table E4-2) x 0.46 (avg. SCLERP from Table E4-3)
| |
| = 1.98E-6/yr.
| |
| The above estimated SLERF will be used for the base case SLERF value for RICT calculations that apply when the primary containment is inerted. If a RICT is being entered during a period when the primary containment is de-inerted, a different SLERF penalty of 4.31E-6/yr.
| |
| (SCDF = SLERF, SCLERP = 1.0) will be applied to address the increased potential for hydrogen deflagration events in the primary containment. This is deemed conservative since the QCNPS Level 2 FPIE PRA (Reference [16]) appropriately models accident progression steam generation inside the primary containment with an estimated 0.5 probability (basic event 1CZPH-STMINRTF--) that the steam inerted condition fails to prevent hydrogen detonation induced failure of primary containment. Therefore, a CLERP of ~0.5 could apply for de-inerted conditions. Given the uncertainty in the steam inerting value of 0.5 and the small timeframe for potential de-inerted conditions, a conservative assumption for SCLERP of 1.0 will apply when the primary containment is de-inerted.
| |
| Application of SLERF in RICT Calculations The SLERF estimate documented above is conservatively used in the RICT process.
| |
| Conservatism in the RICT process derives from the proposed approach to apply the total estimated annual seismic LERF as a delta SLERF in each RICT calculation, regardless of the duration of the completion time. The total estimated annual seismic CDF and LERF will be applied starting at time zero for each RICT calculation.
| |
| E4-22
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models 3.8 Evaluation of Seismic Induced LOOP Past TSTF-505 applications have also included evaluation of any incremental risk associated with challenges to the facility that do not exceed the design capacity and the past submittals have focused on the challenge of seismically induced LOOP. The seismic penalty calculation is intended to address the fraction of seismic events within (i.e., at or below) the design basis by conservatively including very low magnitude seismic events (as low as 0.0005g peak ground acceleration, PGA, i.e., 1/480th of the QCNPS Safe Shutdown Earthquake, SSE of 0.24g PGA) in the SCDF and SLERF convolution calculations. These very low magnitudes are even a factor of 250 lower than the QCNPS Operating Basis Earthquake, OBE of 0.12g PGA; the plant is reasonably expected to remain online for seismic events below the OBE. This appendix provides additional discussions and calculations regarding the inconsequential impact on RICT calculations from plant challenges associated with seismic-induced LOOP from earthquakes within the design basis.
| |
| The approach used in the discussion below is the same as used in past LARs that have explicitly discussed this topic, i.e., 1) estimate the annual frequency of seismic-induced LOOP;
| |
| : 2) assume no offsite AC recovery within 24 hrs.; and 3) compare the result with the internal events PRA frequency estimate for non-recovered LOOP. The methodology used for computing the seismically induced LOOP frequency is to convolve the QCNPS mean seismic hazard curve with an offsite power seismic fragility. Previous TSTF-505 applications have approached this discussion conservatively by performing the seismic-induced LOOP convolution calculation over the entire hazard curve (not just the portion of the hazard curve below the design basis). That same approach is used herein.
| |
| Table E4-4 provides the QCNPS mean PGA seismic hazard data and the LOOP seismic-induced failure probability (increasing with increasing seismic magnitude) based on the seismic fragility of offsite power. The seismic-induced LOOP convolution calculation in Table E4-4 includes the entire seismic hazard curve from earthquakes magnitudes well below the QCNPS operating basis earthquake to well beyond the QCNPS safe shutdown earthquake.
| |
| The failure probabilities for seismic-induced LOOP are represented by failure of ceramic insulators in the offsite AC power distribution system, based on the following seismic fragility data from Table A-0-4 of the NRC RASP Handbook, Volume 2 (Reference [36]). This is a common offsite power seismic fragility used for Central and Eastern US SPRAs and seismic risk calculations:
| |
| Offsite Power Seismic Capacity (ceramic insulators):
| |
| Median Acceleration Capacity, Am = 0.30g PGA Randomness uncertainty, R = 0.30 Modeling uncertainty, U = 0.45 Given the mean frequency and failure probability for each seismic hazard interval, it is straightforward to compute the estimated frequency of seismically induced loss of offsite power for the QCNPS site by multiplying the hazard interval occurrence frequency and the offsite power fragility failure probability. As shown in Table E4-4, the total seismic-induced LOOP frequency across the entire seismic hazard curve is estimated at 2.2E-5/yr. Note that this overstates the "within design basis" challenge frequency but is conservative for this purpose.
| |
| E4-23
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models The QCNPS full-power internal events, FPIE, PRA models LOOP from plant-centered, switchyard-centered, grid-related, and weather-related events. Based on the QCNPS FPIE PRA, the total 24-hr non-recovered LOOP frequency is 1.1E-3/yr., as shown in Table E4-5.
| |
| Assuming offsite AC recovery failure probability of 1.0 for 24 hrs. for seismic-induced LOOP, the total (i.e., across the entire hazard curve) 24-hr non-recovered seismic-induced LOOP frequency is 1.93% of the total 24-hr non-recovered LOOP frequency already addressed in the FPIE PRA. The "within design basis" (i.e., up to the SSE) 24-hr non-recovered seismic-induced LOOP frequency is approximately 1% of the total 24-hr non-recovered LOOP frequency already addressed in the FPIE PRA.
| |
| As can be seen, the 24-hr non-recovered seismic-induced LOOP frequency is a very small percentage of the frequency of such challenges already captured in the FPIE PRA (which is explicitly used in RICT calculations) such that it will not significantly impact the RICT Program calculations, and it can be omitted from explicit analysis in RICT calculations.
| |
| E4-24
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-4: QCNPS Seismic-Induced Loop Frequency Estimate (Across Entire Seismic Hazard Curve)
| |
| Quad Cities Offsite Power Convolution Calculation QC Seismic Hazard Curve (PGA)
| |
| HCLPF (Quad Cities Offsite Power HCLPF fragility with Seismic Hazard)
| |
| Mean Hazard Interval Hazard Hazard Peak Ground Convolved HCLPF Am Exceedance Representative Interval Interval c Acceleration Frequency (g, PGA) (g, PGA) Frequency Magnitude Fragility Occurrence (g) (/yr)
| |
| (/yr) (geo. mean, g PGA) (Mean) Frequency (/yr) 0.09 0.30 0.54 0.0005 5.86E-02 0.0007 1.89E-29 1.98E-02 3.73E-31 0.001 3.88E-02 0.0022 5.52E-20 2.97E-02 1.64E-21 0.005 9.13E-03 0.0071 1.83E-12 4.87E-03 8.93E-15 0.01 4.26E-03 0.0122 1.48E-09 1.80E-03 2.66E-12 0.015 2.46E-03 0.0212 4.37E-07 1.67E-03 7.29E-10 0.03 7.92E-04 0.0387 7.09E-05 4.71E-04 3.34E-08 0.05 3.21E-04 0.0612 1.55E-03 1.64E-04 2.54E-07 0.075 1.57E-04 0.0866 1.02E-02 6.20E-05 6.35E-07 0.1 9.50E-05 0.1225 4.68E-02 4.79E-05 2.24E-06 0.15 4.71E-05 0.2121 2.54E-01 3.38E-05 8.60E-06 0.3 1.33E-05 0.3873 6.75E-01 8.65E-06 5.83E-06 0.5 4.65E-06 0.6124 9.03E-01 2.81E-06 2.54E-06 0.75 1.84E-06 0.8660 9.74E-01 9.44E-07 9.19E-07 1 8.96E-07 1.2247 9.95E-01 6.00E-07 5.97E-07 1.5 2.96E-07 2.1213 1.00E+00 2.63E-07 2.63E-07 3 3.30E-08 3.8730 1.00E+00 2.81E-08 2.81E-08 5 4.87E-09 6.1237 1.00E+00 4.01E-09 4.00E-09 7.5 8.65E-10 8.6603 1.00E+00 6.40E-10 6.40E-10 10 2.25E-10 11.0000 1.00E+00 2.25E-10 2.25E-10 Total Convolved Seismic LOOP Across Hazard Curve (1/yr): I 2.2E-05 E4-25
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-5: Loss of Offsite Power (LOOP) Non-Recovery Frequency QC Cities FPIE PRA QC FPIE PRA QC FPIE PRA LOOP Initiator Probability of 24-hr Non-LOOP Contributor Contributor Non-Recovery of Recovered LOOP Frequency(1) Offsite AC by Frequency (/yr.) (3)
| |
| (/reactor critical yr.) 24 Hrs(2)
| |
| Plant-Centered 2.98E-03 8.90E-03 2.6E-05 Switchyard-Centered 1.11E-02 2.43E-02 2.6E-04 Grid-Related 3.97E-03 2.11E-02 8.1E-05 Weather-Related 3.62E-03 2.22E-01 7.7E-04 Total: 1.1E-03 Ratio of Seismic LOOP to FPIE 24-hr Non-Recovered LOOP: 1.93% (4)
| |
| Table E4-5 Notes:
| |
| : 1. Values per Table D-4 of QCNPS Initiating Events Notebook (QC-PSA-001, Revision 7, June 2019).
| |
| : 2. Values per Table E-2 of QCNPS Initiating Events Notebook (QC-PSA-001, Revision 7, June 2019).
| |
| : 3. Includes QC U1 and U2 average criticality factor of 0.962 per Appendices B and C of QC-PSA-001, Revision 7, June 2019.
| |
| : 4. Ratio: 2.2E-05 / 1.1E-03 = 1.93%.
| |
| 3.9 Summary Estimates of SCDF and SLERF have been derived for use in the QCNPS TSTF-505 program.
| |
| Since the estimates are intended to be treated as conservative values in the RICT calculations for that program, the results (listed below) for the case of plant level fragility HCLPF = 0.24g PGA with c = 0.4 will be used:
| |
| Seismic CDF = 4.31E-6/yr.
| |
| Seismic LERF = 1.98E-6/yr. (inerted containment)
| |
| Seismic LERF = 4.31E-6/yr. (de-inerted containment)
| |
| E4-26
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models 4 EXTREME WINDS ANALYSIS Section 5.2.1 of the IPEEE (Reference [7]) documents the screening of High Winds and Tornadoes; the screening is based on design information from UFSAR Section 3.3 (Reference [37]). Section 3.3.2.1 provides the design parameters for tornado loadings, which include 300 mph tangential velocity, 60 mph translational velocity and a pressure drop of 3 psi in 3 seconds.
| |
| 4.1 Wind Pressure Tornado wind speeds and design pressures bound straight winds for most plant structures, including the exterior concrete walls of the reactor building, turbine building, control room, and steel superstructure of the reactor building and turbine building. However, per Section 3.3.1.1.2 of the UFSAR (Reference [37]), the concrete chimney can only withstand wind pressures up to 217 mph.
| |
| Tornado wind speed hazard curve information for Quad Cities is provided in Table 6-1 of NUREG/CR-4461, Revision 2 (Reference [38]). Based on the EF-Scale, the wind speed for the 1E-6 annual exceedance probability (AEP) is 202 mph. Comparable 1,000,000 year Mean Recurrence Interval (MRI) tornado wind speeds from the ASCE 7 Hazard Tool (Reference [39])
| |
| are less than 200 mph. Therefore, the frequency of the design tornado wind speed for Class I structures is much less than 1E-6/yr and the frequency of winds that could cause the failure of the concrete chimney are also less than 1E-6/yr. Thus, wind pressure effects can be screened using Criterion PS4.
| |
| The ASCE data tornado wind speed data for QCNPS was used to develop the site-specific tornado hazard curve. The ASCE data in Table E4-6 was used to develop the hazard curve which was used for screening wind pressure effects at QCNPS and is also used in the tornado missile risk model, as discussed in Section 4.2, below.
| |
| Table E4-6: ASCE 7 Hazard Tool Tornado Wind Hazard Data for QCNPS Curve Fit(1)
| |
| Wind Speed Exceedance MRI (years) Exceedance (mph) Frequency (/yr)
| |
| Frequency (/yr) 10,000 130 1E-4 1.1E-04 100,000 171 1E-5 1.0E-05 1,000,000 210 1E-6 1.0E-06 10,000,000 248 1E-7 1.2E-07 Notes:
| |
| (1) Curve fit from hazard curve: y = 0.2109e-0.058x (Reference [40])
| |
| 4.2 Tornado Missiles Although Quad Cities conforms to its design basis and licensing basis for tornado missiles, there are several risk significant SSCs (e.g., DG intake and exhaust pipes, certain safety-related E4-27
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models switchgear, and various electrical cables) that are not protected from all tornado missiles.
| |
| Because of this, a conservative tornado missile risk model was developed for each QCNPS unit to assess the risk associated with tornado missiles. The models use the hazard curve developed from the ASCE 7 Hazard Tool data (see Section 4.1), conservative failure probabilities and assumptions (e.g., loss of offsite power with no recovery, guaranteed failure of the SBO DGs).
| |
| The conservative tornado missile risk model was developed from the internal events PRA model and includes the following conservative assumptions:
| |
| : 1. A tornado event is assumed to result in a loss of offsite power (LOOP) with a probability of 1.0 and is considered unrecoverable.
| |
| : 2. Tornado missile failure probabilities for potentially vulnerable SSCs are based on the methodology provided in NEI 17-02 (Reference [41]).
| |
| : 3. SSCs in structures not protected against tornado winds and missiles (e.g., SBO DGs) are assumed to fail with a probability of 1.0. Such SSCs will likely survive at low tornado wind speeds, which comprise the majority of the tornado frequency.
| |
| : 4. No credit taken for FLEX or BlackStarTech equipment.
| |
| Estimates for tornado missile CDF and LERF are shown in Table E4-7. CDF for both units is less than 1E-6/yr and LERF is much less than 1E-7/yr. Unit 2 tornado missile risk is higher, since more Unit 2 risk significant SSCs (primarily switchgear and cables) are potentially vulnerable to tornado missiles, as compared to Unit 1.
| |
| Average maintenance results in Table E4-7 support the screening of tornado missiles; zero maintenance results are provided for perspective and are relevant to the tornado missile penalty factor, discussed below.
| |
| Table E4-7: Tornado Missile Risk Estimates (Reference [40])
| |
| Average Zero Maintenance End State Maintenance (/yr) (/yr)
| |
| U1 CDF 4.8E-07 3.7E-07 U1 LERF 8.1E-09 6.7E-09 U2 CDF 7.2E-07 5.7E-07 U2 LERF 1.6E-08 1.4E-08 Estimates for CDF for both units were less than 1E-6/yr and LERF was estimated to be much less than 1E-7/yr. Therefore, the tornado missile hazard can also be screened using Criterion PS4.
| |
| However, the CDF due to tornado missiles for certain maintenance or LCO configurations is determined to be above 1E-6/yr, requiring a tornado missile (TM) penalty factor to be established for RICT calculations. The QCNPS tornado missile risk assessment (Reference [40]) also documents the calculations used to determine the TM penalty factors E4-28
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models (CDF and LERF). The tornado missile risk model was quantified for all LCO configurations.
| |
| All other maintenance terms were set to FALSE for these configurations.
| |
| For the CDF penalty factor, the highest CDF for Unit 1 is associated with LCO 3.8.7.B/C for DC division inoperability (CDF = 3.9E-6/yr, CDF = 3.5E-6/yr). Many Unit 1 LCO configurations have CDF in the 3E-6 to 4E-6/yr range (e.g., 3.8.1.B for EDGs, 3.5.1.B for two RH Pumps). CDF for other LCOs is all less than 1E-6/yr, including cases with no risk increase (e.g., various instrumentation in 3.3.5.1). The highest Unit 2 CDF is associated with LCO 3.8.1.D for DG 1/2 inoperability (CDF = 5.2E-6/yr, CDF = 4.6E-6/yr). CDF for other LCO configurations is similar to Unit 1 although slightly higher.
| |
| For the LERF penalty factor, the highest LERF for both units is associated with LCO 3.5.1.B for two RH pumps (Unit 1 LERF = 1.9E-7/yr, LERF = 1.8E-7/yr; Unit 2 LERF = 2.6E-7/yr, LERF = 2.4E-7/yr). Several other Unit 2 LCO configurations have LERF in the 2E-7 to 3E-7/yr range (Unit 1 LERF results are all less than 2E-7/yr), but tornado missile LERF for most LCO configurations is less than 1E-7/yr.
| |
| TM penalty factors were conservatively set to provide margin to the highest CDF and LERF for both units. The penalty factors that are established and will be applied to all RICT configurations are:
| |
| CDF 1E-5/yr LERF 5E-7/yr 5 EXTERNAL FLOODING ASSESSMENT 5.1 Plant Siting As reported in the Flood Hazard Reevaluation Report (FHRR) (Reference [42]), QCNPS is located approximately three miles north of the village of Cordova, IL. The plant is located on the Mississippi River at its confluence with Wapsipinicon River.
| |
| 5.2 External Flooding Analysis Evolution 5.2.1 Flood Hazard Reevaluation Report On March 12, 2012, the NRC issued Reference 1 to request information associated with Near-Term Task Force (NTTF) Recommendation 2.1 for Flooding. One of the required responses in this letter directed licensees to submit a FHRR, including the interim action plan (to address calculated higher flooding hazards relative to the design basis prior to completion of the NTTF flooding integrated assessment) requested in Item 1.d of Reference 1, Enclosure 2, if appropriate. QCNPS submitted the FHRR on March 12, 2013, to NRC for review (Reference [42]). This report included analysis in accordance with applicable present-day guidance at the time it was submitted and represents the most recent deterministic flood hazard analysis for the site. The FHRR contains the analysis results for ten (10) flooding cases (eight flood causing mechanisms, one combined mechanism case, and one associated effects case),
| |
| E4-29
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models including screening justification and probable maximum flood elevations. The final output from the QCNPS FHRR is the comparison of the QCNPS current design basis (CDB) flood elevations and the reevaluated flood elevations. Any flood causing mechanisms that are not bounded by the CDB must be included in additional analysis.
| |
| Table E4-8 below contains the results of the QCNPS CDB comparison to the reevaluated flood hazard from the QCNPS FHRR (Reference [42]). The QCNPS FHRR determined that two flood causing mechanisms and one associated effect are not bounded by the QCNPS CDB:
| |
| : 1. Local Intense Precipitation
| |
| : 2. Combined Effects (Probable Maximum Flood + Dam Failure + Wind-Generated Waves)
| |
| : 3. Hydrodynamic and Debris Loading Table E4-8: Flood Causing Mechanisms Summary Results Flood Hazard Flood-Causing QCNPS CDB Reevaluation Comparison Mechanism Elevation Local Intense Precipitation Not Considered Varies Not Bounded Flooding in Streams and 603 feet MSL 600.5 feet MSL Bounded Rivers (PMF)
| |
| Dam Breaches and Not Considered 600.9 feet MSL Bounded Failures Storm Surge Not Applicable Not Applicable Not Applicable Seiche Not Applicable Not Applicable Not Applicable Tsunami Not Applicable Not Applicable Not Applicable Ice-Induced Flooding Not Considered 579.8 feet MSL Bounded Channel Migration or See Note 1 See Note 1 Bounded Diversion Combined Effects (PMF + Not Considered 605.0 feet MSL Not Bounded Dam Failure + Wind-Generated Waves)
| |
| Hydrodynamic and Debris Not Considered See Note 2 Not Bounded Loading Table E4-8 Notes:
| |
| : 1. Channel migration or diversion is not an issue because the river flow and geometry is controlled by United States Army Corp of Engineers (USACE) navigational structures.
| |
| : 2. Impacts from hydrodynamic and debris loads will be further evaluated in the Integrated Assessment.
| |
| The NRC issued a staff assessment of the FHRR results on November 18, 2016 (Reference [43]). The staff concluded that the results in the FHRR were appropriately evaluated E4-30
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models and should serve as input to the Integrated Assessment (IA). Therefore, the FHRR represents the most up-to-date flooding analysis and will also serve as the input to the IA.
| |
| 5.2.2 Integrated Assessment Following the completion of the FHRR, it was determined that the three flood causing mechanisms that were not bounded by the CDB need to be evaluated for their impact to the plant in an IA (Reference [44]). The guidance for performing an integrated assessment was provided in NEI 16-05, Revision 1 (Reference [45]).
| |
| NEI 16 05 indicates that each flood-causing mechanism not bounded by the Design Basis (DB) flood (using only stillwater and/or wind-wave runup level) should follow one of the following five assessment paths:
| |
| Path 1: Demonstrate Flood Mechanism is Bounded Through Improved Realism Path 2: Demonstrate Effective Flood Protection Path 3: Demonstrate a Feasible Response to Local Intense Precipitation (LIP)
| |
| Path 4: Demonstrate Effective Mitigation Path 5: Scenario Based Approach Non-bounded flood-causing mechanisms in Paths 1, 2, or 3 would only require a Focused Evaluation to complete the actions related to external flooding required by the March 12, 2012, 10 CFR 50.54(f) letter. Mechanisms in Paths 4 or 5 require an IA.
| |
| The reevaluated flood hazard, summarized in Reference [43], was utilized as input to the Flooding IA. The Flooding IA reaffirms that QCNPS has appropriately addressed plant vulnerabilities to external flooding and will not require additional safety enhancements since mitigating strategies (FLEX) remain feasible (per Reference [43]) for the reevaluated flood hazard and effective protection of Key Safety Functions (KSF) has been demonstrated for higher likelihood flooding scenarios.
| |
| There are two components to the IA utilized by QCNPS in the assessment. The first is the determination that the site has adequate protection from the LIP event and is achieved by closing six (6) flood tight gates prior to the arrival of the storm. These gates protect the plant, with margin, from the maximum calculated flood depth from extreme rainfall. This path, known as Path 2, demonstrates physical protection from a flooding event.
| |
| The second component includes characterizing the flooding scenario into two bins, scenario 1 and scenario 2. The first scenario bin is high likelihood flood events that are calculated to be greater than 1E-4/yr. The second scenario bin includes lower likelihood events that are less than 1E-4/yr. For scenario 1, adequate protection (physical) must be shown to show that there is low likelihood of impact to the site from more frequent storm events. For scenario 2, feasible mitigation can be utilized for extreme storm events that have a low likelihood of occurrence.
| |
| The hazard frequency was developed utilizing historical data from stream gauges and paleo flood information to derive a frequency for storm elevations more frequent than 1E-4/yr. This E4-31
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models analysis is outlined in the IA. The tables below show the flood scenario bins developed and the resulting frequency for each scenario bin. QCNPS ground floor elevation of critical structures is 595.0 feet MSL. Below plant grade, flooding will not result in any impacts to the site. Flooding above ground floor elevation 595.0 feet MSL, will follow the site's feasible mitigation strategy, as outlined in the UFSAR (Reference [37]).
| |
| Table E4-9: Description of Likelihood and Flood Strategy Flood Critical Water Surface Likelihood Description of Flood Strategy Scenario Elevation (feet MSL 1912)
| |
| Ground Floor Elevation2 -
| |
| 1 595.0 High Effective Protection 2 > 595.0 to 605.0 Low Feasible Response/Mitigation Table E4-10: Available Physical Margin Annual Exceedance Water Surface Elevation Available Physical Margin to Ground Probability (feet MSL 1912) Floor El2 595.0 (feet MSL 1912) 10 Median
| |
| -4 591.0 4.0 feet 10-3 @ 95%
| |
| 591.5 3.5 feet Confidence Limit 10-3 Median 589.2 5.8 feet The conclusion of the Integrated assessment was that Quad Cities has adequate protection for the LIP mechanism and more frequent flooding events due to combined effects. The station also demonstrated a feasible mitigation strategy for less frequent flood (those with a frequency lower than 1E-4/yr).
| |
| NRC issued an assessment of the IA on August 29, 2019 (Reference [46]). The NRC agreed with the finding in the IA stating that QCNPS has reliable protection measures against LIP and the Time Sensitive Actions (manual actions) were determined feasible.
| |
| For the combined effects flood, the staff concluded that for high likelihood floods, the station has effective flood protection, that there is available physical margin, that its effective flood protection is reliable, and that it does not rely on human actions.
| |
| 2 The terminology used in the IA for elevation 595.0 is plant grade or site grade. However, the UFSAR states that 594.5 is plant grade or site grade and 595.0 is ground floor elevation. Since 595.0 equates to both ground floor elevation and the elevation where water can begin to enter the plant, the text has been revised to remove references to site grade at 595 and replace with ground floor elevation.
| |
| E4-32
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models For combined effects low likelihood flooding, the staff concluded as follows:
| |
| The NRC concludes that, based on the information provided by the licensee in the IA, the licensee has a feasible approach for addressing a flood of 600.9 ft MSL stillwater elevation with 4.1 ft. of wave runup. The licensee has shown that the mitigation strategies related to the combined events flood causing mechanism (in a similar manner to the reevaluated LIP hazard) does not rely on the use of mitigation strategies equipment, and instead relies on normal shutdown and licensing basis decay heat removal procedures. The staff has confirmed that no changes are needed to the COB shutdown strategy, that the strategies within buildings, including connection points, are not impacted by the reevaluated hazard, and that the dynamic effects from the reevaluated hazards on the outside portion of key structures will have negligible impact on the station's ability to reach and maintain safe shutdown. (Reference [46])
| |
| 5.2.3 External Flooding Screening Evaluation 5.2.3.1 Screening Mechanisms Based on FHRR and IA Results This section summarizes the external flooding mechanisms that do not pose a challenge to the QCNPS site or the design of the site bounds the impacts for the hazard. The input for this evaluation can be found in Enclosure 2 of the QCNPS FHRR report (Reference [42]) and is confirmed by NRC in Table 3.1-2 in their assessment of the QCNPS FHRR (Reference [43]).
| |
| Table E4-11: Flooding Mechanisms Screened in FHRR and IA Flooding Hazard Screening Reason Reference Mechanism Criteria(1)
| |
| Storm Surge EXT-B1: 3 Not plausible at the site FHRR Enclosure 2 Table 1 (Reference [42])
| |
| Seiche EXT-B1: 3 Not plausible at the site FHRR Enclosure 2 Table 1 (Reference [42])
| |
| Tsunami EXT-B1: 3 Not plausible at the site FHRR Enclosure 2 Table 1 (Reference [42])
| |
| Channel EXT-B1: 1 As indicated in the FHRR Enclosure 2 Migrations / UFSAR - The authority Table 1 (Reference [42])
| |
| Diversions to control the river is vested in the USACE.
| |
| Should the need to control the river arise, Exelon will make the required notification to the USACE. These arrangements have been made and are detailed in the QCNPS emergency procedures.
| |
| E4-33
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-11: Flooding Mechanisms Screened in FHRR and IA Flooding Hazard Screening Reason Reference Mechanism Criteria(1)
| |
| Hydrodynamic EXT-B1: 1 The LIP barriers and FHRR Enclosure 2 and Debris power block structures Table 1 (Reference [42])
| |
| Loading are designed to withstand the max IA Table 3 hydrodynamic (3.6 lb/ft) (Reference [44])
| |
| and debris loads (480 lbs). Table E4-12 lists the barriers credited and their capacity ratings with respect to the expected loads.
| |
| Table E4-11 Notes:
| |
| : 1. These hazard screening criteria are outlined in Part 6, Supporting Requirement EXT-B1, of Reference [3].
| |
| The first three mechanisms listed in Table E4-11 above are not plausible at the QCNPS site due to the location of the station in the middle of the continental United States. Channel Migrations /
| |
| Diversions and Hydrodynamic / Debris Loading were identified as being bounded by the CDB of the plant.
| |
| 5.2.3.2 Local Intense Precipitation Current Risk Basis The local intense precipitation (LIP) event was not included in the plant original design basis.
| |
| The LIP was reevaluated in the FHRR (Reference [42]) and found that extreme precipitation could cause flooding above ground floor elevation (595 ft). QCNPS has 13 credited flood barriers at key openings around the power block to keep LIP flood waters from entering the buildings during the event as shown in Figure E4-2. Six of the barriers are temporary and passive requiring manual installation (as indicated in the Requires Manual Closure column of Table E4-12). The remaining barriers are permanently installed exterior doors or plates that do not require manual actions to perform their function. All exterior credited flood barriers, their design, and available physical margin (APM) were evaluated in the Integrated Assessment report and documented in calculation QDC-0000-S-2089, Revision 2 (Reference [47]). The installation of the temporary barriers is governed by QCOA 0010-22, "Local Intense Precipitation Response Procedure" (Reference [48]), which includes rainfall monitoring guidance and action triggers. QCOA 0010-22 provides rainfall symptoms that trigger commencement of Enhanced Monitoring using specified forecast sources.
| |
| E4-34
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-12: QCNPS LIP Flood Barriers Margin in Margin in Barrier Requires LIP Flood Maximum Barrier Flood Depth Barrier Description Height Manual Flooding Depth per Structural Number Due to Over-
| |
| [ft] Closure Depth [ft] Capacity IC [ft]2,4 Topping [ft]
| |
| [ft]
| |
| 1A/1B U1 & U2 Turbine Building Roll Up Doors 4.5 Yes 3.52 0.98 0.79 0.94 (2 Barriers) 2A/2B U1 & U2 Turbine Building/Reactor Building Interlock to HRSS Personnel N/A(5) No 3.50 N/A 0.48 0.68 Doors 3 Reactor Building 1/2 Trackway Door N/A(5) No 3.56 N/A 0.44 0.78 4 1/2 EDG Interlock Door N/A(5) No 3.62 N/A 0.77 0.78 5 Steel Plate on North Side of 1/2 EDG N/A(5) No 3.58 N/A 0.47 0.71 Building 6 Unit 2 Turbine Building Northwest Personnel N/A(5) No 1.07 N/A 0.40 0.998 Access Door 7 Turbine Building to Radwaste Building N/A No N/A N/A N/A N/A Door 8 LTD Building to Trackway 1 Door 2.5 Yes 1.28 1.22 0.98 0.94 9 Personnel Decon Room Door 4.0 Yes 3.52(3) 0.48 0.48 0.998 10 Service Building to Trackway 1 Door 4.0 Yes 3.52(3) 0.48 0.48 0.40 11 Aux Electric Room Door 4.0 Yes 3.52(3) 0.48 0.48 0.85 TB North Turbine Building North Siding(1)
| |
| N/A(5) No 1.89 N/A N/A 0.53 Siding E4-35
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-12 Notes (Information excerpted from Table 8.1A of QDC-0000-S-2089 (Reference [47]):
| |
| : 1. The TB Siding is evaluated to the maximum flood elevation without margin.
| |
| : 2. Structural ICs ("Interaction Coefficients") consider flooding depth with corresponding margin and associated hydrodynamic loads per References 40/44 of QDC0000-S-2089 (Reference [47]) when considering faulted allowables.
| |
| : 3. Conservatively the maximum flood water elevation for the entire TB/SB is used.
| |
| : 4. Structural IC is the ratio of the LIP loading to the design loading. A Structural IC of less than 1.0 indicates that the LIP load is less than the design load and, therefore, is considered acceptable. The lower the IC, the greater the structural margin.
| |
| : 5. N/A in this column indicates that the flood barrier does not have a "height" as they are doors or siding that extend well beyond the maximum flood height. These barriers are only credited against their structural IC, shown in the last column.
| |
| E4-36
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Analysis No. QOC-0000-5-20B9 Revision No. 2 Page No.19 J
| |
| UNIT l UNIT 2 A
| |
| - --© o, E,
| |
| 8 Figure E4-2: LIP Barrier Locations E4-37
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Weather monitoring is directed in Procedure QCOA 0010-22. Section A of this procedure provides multiple separate triggers which, when met, initiate implementation of the main body of the procedure.
| |
| If at any time, the short-term forecast predicts 3-inches or more of precipitation in a 1-hour period, then the following actions will be completed within an hour (Steps D.1):
| |
| Equipment operators are dispatched to clear any blocked storm drains if conditions allow.
| |
| Equipment operators are dispatched to install temporary flood barriers, including the Unit 1 & Unit 2 Turbine Buildings Roll Up Doors Fastlogs, LTD Building to Trackway 1 LIP barrier and Personnel Decon Room Fastlogs; Equipment operators verify that the LIP Barrier Fastlogs are installed on the FLEX building; Equipment operators verify that external doors (listed individually within the procedure) relied upon for protection are closed; Equipment operators are dispatched to key site locations to provide equipment status reports directly to the Control Room.
| |
| The QCNPS site response to LIP was evaluated for adequacy in Section 7 of the IA (Reference [44]). A complete list of actions, credited flood barriers, and evaluation of their ability to protect reliable flood protection was provided. Table E4-12 lists the flood barriers utilized in the QCNPS LIP protection strategy, their margin to overtopping or structural capacity, and if they require manual action to install.
| |
| Table E4-12 shows that there are six barriers (denoted in the column "Requires Manual Closure" with a value of "Yes") that require manual closure prior to the arrival of flood waters.
| |
| These barriers are temporary, fast installing, and credited with screening the flooding scenario.
| |
| The remaining credited barriers do not require manual action, considered permanent, and passive protection with adequate available physical margin (APM).
| |
| The flood barrier installation feasibility assessment is documented in Section 7.1.3 of the QCNPS IA (Reference [44]). The following elements are included in the QCNPS IA justification for an adequate site response:
| |
| : 1. Defining Critical Path and Time Sensitive Actions (TSAs)
| |
| : 2. Demonstrating all TSAs are Feasible
| |
| : 3. Establishing Unambiguous Procedural Triggers
| |
| : 4. Proceduralized and Clear Organizational Response to Flooding
| |
| : 5. Detailed Flood Response Timeline In summary, the TSAs considered are the installation of the six barriers that require manual actions shown in Table E4-12. The longest barrier installation takes approximately 45 minutes and all six can be done concurrently. Validation Plan #12 (Reference [49]) confirms the barriers can be installed with eight equipment operators in approximately one hour and strain on these operators is minimal. Operator training is provided for installation of the flood barriers in E4-38
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models LN-FLEX.2, Section J (Reference [50]). Given that the timeline is well defined, the entry conditions are laid out in the procedures, and training is provided, QCNPS has demonstrated that the site response to LIP is adequate. This was confirmed by NRC in the assessment of the IA (Reference [46]).
| |
| RICT Disposition for LIP The QCNPS IA confirmed that the site has adequate protection from the LIP flooding mechanism. There are no direct external flooding impacts to the plant once the six LIP barriers are installed. Therefore, this mechanism can be screened from further consideration utilizing the initial preliminary screening EXT-B1 Criterion 1, "Event damage potential is less than events for which the plant is designed."
| |
| There are no configuration specific considerations for screening this hazard.
| |
| 5.2.3.3 Combined Effects (PMF + Dam Failure + Wind-Generated Waves)
| |
| Current Risk Basis As stated above, QCNPS site external flooding from streams and rivers, dam failure and combined effects PMF were reevaluated in the FHRR (Reference [42]). It was found that when combining a PMF with hydrologic upstream dam failure and wind/wave action, the reevaluated flood levels were not bounded by the CLB of the plant. Therefore, an IA (Reference [44]) was performed to determine the adequacy of the QCNPS site response to extreme external flooding.
| |
| The combined effects flood was broken into two scenarios, one of high likelihood (defined by floods with an exceedance frequency of greater than 1E-4/yr) and those of low likelihood (less than 1E-4/yr). The frequency analysis utilized in the IA was performed using historical records and paleo flood data to develop the estimates used in this scenario characterization. Details regarding the development of the hazard frequencies can be found in Section 7.2.1 of the QCNPS IA (Reference [44]). The conclusion was that for high likelihood floods, there is adequate protection due to the flood events remaining below ground floor elevation and no impacts to plant SSCs responsible for maintaining KSF were anticipated from this flooding scenario. For low likelihood events, the CLB flood mitigation strategy (documented in the UFSAR (Reference [37]) provided an adequate site response to extreme flooding events. The current flood mitigation strategy is governed by procedure QCOA 0010-16 (Reference [51]).
| |
| Hazard Curve for PMF To further enhance the understanding of risk to the site from the combined effects PMF mechanism, a probabilistic flood hazard assessment was performed in 2021 for QCNPS (Reference [52]). This study utilized current day practices to develop a stochastic weather model that served as input to a continuous hydrologic model. The output from the study included a stage-mean frequency table characterizing floods with a frequency as low as 1E-6/yr.
| |
| This table was utilized to plot the curve shown in Figure E4-3 and serves as a key input to the following risk analysis.
| |
| E4-39
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models WATER SURFACE ELEVAT ION VS. ANNUAL EXCEEDANCE PROBABILITY 599 598 rbp of Flood Bar iers - 599 .0' << lE-6/YR 597 596 l l I11 595 Plant Grade - 59 .O' ~ 2E-6/YR 594 593 592 t
| |
| 591 f
| |
| 590 -
| |
| z 589 0
| |
| ~
| |
| 588 LIJ
| |
| ..J 587 LIJ t
| |
| LIJ u
| |
| 586 ~
| |
| a:
| |
| 585 ::i Vl a:
| |
| 584 I
| |
| LIJ
| |
| ~
| |
| 583 3 I
| |
| 582 581 580 J 579 578 577 l.00E-01 1 .00E-02 l.00E-03 l.00E-04 l.00E-05 l.00E-06 ANNUAL EXCEEDANCE PROBABILITY
| |
| .,... Mean XF Hazard Curve Figure E4-3: External Flooding Hazard Curve for Combined Effects Flooding E4-40
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models LIP Flood Barrier Upgrades and Deployment Following the completion of the PFHA, additional risk reduction measures were evaluated by the station to further reduce the overall risk from the combined effects flood mechanism. It was determined that with minor modifications, the LIP barriers could be deployed well in advance of flood waters arriving from a combined effects flood and protect the station from floods with a water surface elevation (WSE) of 599.0 or lower. Utilizing the information from the hazard curve, it was determined that LIP barrier modification could reduce the risk from the flood mechanism by more than an order of magnitude. Therefore, the station prepared ECs 636912 and 636914 (References [53] and [54]) where all LIP barriers will be modified to protect the plant up to 599.0. The main modification raises one of the barriers from 2.5 tall to 4 tall. The functionality, design and implementation of the barriers remain unchanged. Completion of EC 636914 related modifications is tracked in Attachment 5 as a RICT Program implementation item. Additional verifications were also performed to ensure that all credited flood boundary elements (e.g., building walls, exterior doors, etc.) have adequate design margin to withstand the hydrostatic and hydrodynamic forces from the additional water accumulation around the plant.
| |
| The procedure QCOA 0010-16 was reviewed and changes were identified to include deployment of the six temporary LIP barriers prior to flood waters arriving on site that are consistent with the procedure steps in QCOA 0010-22. Given that QCOA 0010-16 has specified a 96-hour warning time, the timing and validation of deployment is considered subsumed by the Validation #12 (Reference [55]) performed for deployment during a LIP event which has less than a 6-hour warning time in QCOA 0010-22 and the validation exercise remains appropriate for deployment of the barriers in QCOA 0010-16. Revisions to QCOA 0010-16 are planned to be completed prior to the implementation of the RICT and 50.69 programs, in conjunction with the completion of the barrier modification.
| |
| It should be noted that none of the changes proposed to the mitigation strategy for the combined effect flood will impact the CLB strategy outlined in the UFSAR (Reference [37]). For floods that exceed 599.0 and overtop the barriers, the UFSAR strategy will still be implemented and available to mitigate the higher magnitude, low frequency floods. This strategy is not credited in the screening of the hazard.
| |
| Conservative CDF Evaluation The evaluation of a conservative CDF for QCNPS due to the combined effects flooding from a PMF, upstream dam failure and wind/wave action includes two parts. Understanding the mean hazard frequency and development of a conservative analysis of the reliability of the flood mitigation strategy.
| |
| External Flood Scenarios for the Combined Effect Flood From Figure E4-3, the frequency of the combined effects flood exceeding plant grade is approximately 2E-6/yr and the frequency of external floods capable of overtopping the LIP flood barriers is much less than 1E-6/yr with considerable margin. Therefore, the frequency associated with Scenario 2 and the overtopping of the LIP Barriers can be assumed negligible for the purposes of this calculation as the frequency is more than an order of magnitude lower than Scenario 1. Table E4-13 below lists the two external flooding scenarios used in this conservative CDF evaluation.
| |
| E4-41
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-13: : External Flood Scenarios for Combined Effects Flood External Flood Scenario Description Exceedance Frequency Scenario 1 Floods greater than 595.0 2E-6/yr Scenario 2 Floods greater than 599.0 (negligible)
| |
| Analysis of the Mitigation Strategy Similar to the LIP response, the station relies on receiving warning of an impending flood.
| |
| Floods capable of reaching plant grade or higher at QCNPS require a very large storm upstream in the basin. This rainfall is conveyed downstream through the watershed until it arrives at the site. The procedure includes triggers in QCOA 0010-16 for starting flood mitigation preparations. Operators will begin executing the steps in the procedure under two conditions:
| |
| A.1 Flood forecast by the U.S. Weather Bureau A.2 River Level is > 582 Subsequent monitoring is performed to direct operations to shut down the reactors under the following two conditions in Step D:
| |
| Immediately when actual river level exceed 586 as monitored by the gauge painted on the log boom support/north intake bay wall at the plant intake bay Water level predicted to be > 594 mean sea level USGS datum in < 96 hours (The actions of this procedure are intended to be completed in < 72 hours, providing 24 hours of margin).
| |
| QCOA 0010-16 will then direct operators to carry out the remaining steps in the CLB mitigation strategy to ensure extreme floods in Scenario 2 have a mitigation strategy. For this analysis, we will conservatively assume that following the completion of actions in the CLB strategy (approximately 72 hours after shutdown) the procedure will direct the installation of the LIP barriers utilizing the adapted procedural steps and guidance as outlined in QCOA 0010-22.
| |
| This results in a 24-hour system window for installation of the LIP barriers. The procedural triggers are unambiguous and allow the site ample time to initiate the procedure. Given that the current system window for installation of the LIP barriers in QCOA 0010-22 is 6 hours after receipt of major storm warning, the additional time for the combined effects flood can be judged to only increase the reliability of the actions required to install the LIP barriers before the river level rises above grade.
| |
| As stated above, the IA is based on NRC endorsed guidance provided in NEI 16-05 (Reference [45]). Appendix C of Reference NEI 16-05 provides criteria for evaluating flood mitigation strategies based largely on the human reliability analysis performed for PRAs. The deployment of the LIP Barriers was conducted utilizing the guidance in NEI 16-05 and documented in the IA (Reference [44]). The following criteria were evaluated to demonstrate the site had and adequate response to flooding and deploying the LIP Barriers.
| |
| E4-42
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Overall Evaluation of the Flood Response Strategy Defining Critical Path and Time Sensitive Actions (TSAs)
| |
| Demonstrating All TSAs are Feasible Establishing Unambiguous Procedural Triggers Proceduralized and Clear Organizational Response Detailed Flood Response Timeline Accounting for Expected Environmental Conditions NRC issued a staff assessment of the findings in the IA (Reference [46]), agreeing with the conclusions that all the above criteria was evaluated appropriately, and the site has an adequate mitigation strategy in place for LIP barrier installation and the combined effects river flood.
| |
| The site has in place training for installation of the LIP Barriers in LN-FLEX.2, Section J (Reference [50]). A validation (#12) exercise was carried out to demonstrate the strategy takes less than an hour to complete (all barriers being installed concurrently) or 3 hours if performed in series, and staffing requirements were reviewed to ensure that 8 EOs are available to perform the procedure steps in accordance with the minimum staffing requirements. In addition, the conditions under which the operators are to install the LIP barriers are not anticipated to be adverse. The barriers will be installed well in advance of flood waters arriving, with nominal environmental factors (the storm required to induce this level of flooding must happen well upstream in the basin), and with normal offsite power available until the river reaches 595.0.
| |
| Lastly, the site operators and personnel are well trained on both deployment of the LIP barriers and the CLB river flood strategy. QCOA 0010-16 is well laid out with guidance on performing the tasks necessary for successful completion. The procedure does require modifications to install the LIP barriers during a river flood, however, QCOA 0010-22 is an established procedure that site personnel get training on already. The addition of the LIP barrier installation to QCOA 0010-16 with the available time will greatly reduce the risk to the station from external river floods.
| |
| Conservative Reliability Assessment of LIP Barrier Installation The analysis of the mitigation strategy above and the presentation of the strategies to NRC through the IA submittal demonstrates a reliable set of actions used to mitigate CLB external flood causing mechanisms. It has been determined that the time margin available to install the LIP barriers is very large, the environmental factors are expected to be nominal, and the operators receive training on the installation of the LIP barriers. This analysis will utilize a conservative approach to estimating the human error probability (HEP) of installing the barriers.
| |
| The most complex barrier install with the longest installation time (Barriers 1A and 1B) will be used for installing each of the six barriers. This is a conservative assumption because Barriers 3, 8, 9 and 11 are swing gates that require much less time to install.
| |
| E4-43
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-14: External Flood HEPs for LIP Barrier Installation HEP Name Barrier HEP Value HEP-1A 1A 5E-2 HEP-1B 1B 5E-2 HEP-3 3 5E-2 HEP-8 8 5E-2 HEP-9 9 5E-2 HEP-11 11 5E-2 XF-LIP-HEP All 3E-1 From the six HEPs, a single combined HEP can be developed by summing the individual HEPs.
| |
| Should one barrier fail to be installed, the entire boundary fails, and the site would rely on the CLB strategy. However, this analysis has not developed a reliability analysis for the CLB strategy and credit for the mitigation capability will be conservatively ignored.
| |
| CDF Estimate The risk from both scenarios has been evaluated in Table E4-15 below. Scenario 1, where flood waters are greater than 595.0 but lower than 599.0, has a CCDP equal to the combined LIP Barrier installation HEP and conservatively ignoring any credit for the CLB mitigation strategy. In Scenario 1 it is assumed if a single barrier fails (XF-LIP-HEP), the scenario results in core damage. Scenario 2 similarly does not take credit for the CLB mitigation strategy and very conservatively assumes core damage once the flood waters reach the top of the LIP barriers. Given the low frequency of the scenario, however, the contribution from Scenario 2 can be ignored given that the frequency of the scenario alone is at least an order of magnitude lower than that of Scenario 1.
| |
| The summation of both scenarios CDF estimate represents the total bounding mean estimate of CDF for the combined effects flooding mechanism.
| |
| Table E4-15: QCNPS CDF Estimate from Combined Effects Floods QCNPS Site Occurrence CCDP CDF External Flooding Scenario Frequency Scenario 1: Floods greater than 595.0 1.95E-6/yr 0.30 6E-7 Scenario 2: Floods greater than 599.0 1 Total CDF Estimate: 6E-7 This estimate of CDF represents a conservative result due to a very high CCDP for installation of the barriers and their reliability to perform their design function. No additional credit has been taken for defense in depth measure afforded by FLEX pumps, temporary dams and the station will be in a shutdown condition well before the flood waters arrive. The evaluation of these two scenarios with the assumptions made constitute a conservative analysis with margin to the CDF screening limit of 1E-6/yr as described inEXT-C1 Criterion B (Reference [3]).
| |
| E4-44
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models RICT Disposition for Combined Effects Floods The evaluation of CDF from the combined effects flood was conservatively calculated to be 6E-7/yr. Given the conservativisms present in the assumptions, the combined effects flood can be screened utilizing EXT-C1 Criterion B where the bounding mean CDF is less than 1E-6/yr with considerable margin.
| |
| There are no RICT configuration specific considerations that impact this external flooding screening calculation.
| |
| 5.2.4 Conclusion In conclusion, there are only two external flood causing mechanisms that pose a challenge to the site that required screening in this analysis: 1) LIP and 2) the combination of PMF, upstream hydrologic dam failure, and coincident wind/wave action (known as combined effects floods). The LIP flood waters are mitigated after six temporary barriers are installed and the exterior doors are in their normal (closed) position around the power block reliably preventing water from entering the buildings. Once the barriers are installed, there is no challenge to the stations KSFs during this flood event. This mechanism can be screened using the initial preliminary screening EXT-B1 criterion 1 where "the event damage potential is less than events for which the plant is designed."
| |
| For the combined effects floods, a PFHA was performed to characterize the frequency of floods exceeding plant grade and the LIP barriers. Following minor revisions to QCOA 0010-16, the LIP barriers will be installed prior to the arrival of flood waters protecting the plant up to an elevation of 599.0 feet which equates to an exceedance frequency much less than 1E-6/yr and can be assumed negligible. The frequency of water exceeding the ground floor elevation of 595.0 feet is approximately 2E-6/yr, as shown in the hazard curve (Figure E4-3). Failure to install all six LIP gates prior to river level exceeding 595.0 feet is conservatively assumed to have a failure probability of 0.3. All other mitigation capabilities, such as, the CLB strategy/FLEX, temporary dams, and that the plant will be in a shutdown condition, are conservatively ignored leading to the CCDP to be estimated at 0.3, which is equal to the HEP of installing all the gates with no credit for additional mitigation capability.
| |
| When combining the CCDP with the frequency for the two scenarios of flood waters exceeding plant grade and then topping the LIP barriers, the bounding mean CDF can be estimated at 6E-7/yr. This estimate has margin from the screening limit of 1E-6/yr utilizing many conservative assumptions of the plants mitigation capabilities that provide additional confidence in the estimate used for processive screening under EXT-C1 criterion B.
| |
| 6 EVALUATION OF OTHER EXTERNAL EVENT CHALLENGES AND IPEEE UPDATE RESULTS This section provides an evaluation of other external hazards. The results of the assessment of these hazards are provided in Table E4-16. Table E4-17 provides the summary criteria for screening of the hazards listed in Table E4-16.
| |
| E4-45
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Hazard Screening The IPEEE for QCNPS provides an assessment of the risk to QCNPS associated with these hazards. Additional analyses have been performed since the IPEEE to provide updated risk assessments of various hazards, such as aircraft impacts, industrial facilities and pipelines, and external flooding. These analyses are documented in the UFSAR (Reference [37]). Table E4-16 reviews and provides the bases for the screening of external hazards, identifies any challenges posed, and identifies any additional treatment of these challenges, if required. The conclusions of the assessment, as documented in Table E4-16, assure that the hazard either does not present a design-basis challenge to QCNPS, or is adequately addressed in the PRA.
| |
| In the application of RICT, a significant consideration in the screening of external hazards is whether particular plant configurations could impact the decision on whether a particular hazard that screens under the normal plant configuration and the base risk profile would still screen given the particular configuration. The external hazards screening evaluation for QCNPS has been performed accounting for such configuration-specific impacts. The process involves several steps.
| |
| As a first step in this screening process, hazards that screen for one or more of the following criteria (as defined in Table E4-17) still screen regardless of the configuration, as these criteria are not dependent on the plant configuration.
| |
| The occurrence of the event is of sufficiently low frequency that its impact on plant risk does not appreciably impact CDF or LERF (Criterion C2)
| |
| The event cannot occur close enough to the plant to affect it (Criterion C3)
| |
| The event which subsumes the external hazard is still applicable and bounds the hazard for other configurations (Criterion C4)
| |
| The event develops slowly, allowing adequate time to eliminate or mitigate the hazard or its impact on the plant (Criterion C5)
| |
| The next step in the screening process is to consider the remaining hazards (i.e., those not screened per the above criteria) to consider the impact of the hazard on the plant given particular configurations for which a RICT is allowed. For hazards for which the ability to achieve safe shutdown may be impacted by one or more such plant configurations, the impact of the hazard to particular SSCs is assessed and a basis for the screening decision applicable to configurations impacting those SSCs is provided.
| |
| As noted above, the configurations to be evaluated are those involving unavailable SSCs whose LCOs are included in the RICT program.
| |
| E4-46
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| An aircraft (either a portion of Acceptance criterion 1.A of Standard Review Plan 3.5.1.6 (e.g., missile) or the entire (Reference [56]) states the probability is considered less than an aircraft) that collides either order of magnitude of 10-7 per year by inspection if the plant-to-directly or indirectly (i.e., skidding airport distance D is between 5 and 10 statute miles, and the impact with one or more projected annual number of operations is less than 500 D2, or the structures, systems, or plant-to-airport distance D is greater than 10 statute miles, and the components (SSCs) at or in the projected annual number of operations is less than 1000 D2 (PS2, plants analyzed area causing PS4).
| |
| functional failure.
| |
| Per the IPEEE (Reference [7]), the closest airport to the plant is the Secondary hazards resulting Clinton Municipal Airport, a small, public, general aviation facility PS2 from an aircraft impact include, located approximately 9 miles north-northwest of the plant. Airport Aircraft Impact but are not necessarily limited to, data (Reference [57]) shows the annual options from this airport is PS4 fire. less than 12,000 which is less than 500 D2 (PS2, PS4).
| |
| Quad Cities International Airport, about 21 miles southwest of the plant, is the nearest airport with scheduled commercial air service.
| |
| Airport data (Reference [58]) shows the annual operations from this airport to be less than 29,000 which is less than the 1000 D2 criteria (PS2, PS4).
| |
| Based on this review, the aircraft impact hazard is considered negligible.
| |
| 3 The list of hazards and their potential impacts considered those items listed in Tables D-1 and D-2 in Appendix D of RG 1.200, Revision 3 (Reference [71]).
| |
| E4-47
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| Rapid flow of a large mass of The mid-western location of the plant precludes the possibility of an accumulated frozen precipitation avalanche.
| |
| and other debris down a sloped surface resulting in dynamic Based on this review, the avalanche hazard can be considered Avalanche loading of SSCs at or in the C3 negligible.
| |
| plants analyzed area causing functional failure or adverse There are no configuration-specific considerations for this hazard.
| |
| impact on natural water supplies This hazard can be excluded from the TSTF-505 program used for heat rejection. evaluation.
| |
| Accumulation or deposition of Per UFSAR Section 9.2.1.4 (Reference [37]), piping and heat vegetation or organisms (e.g., exchanger intrusion by bivalves (such as Asiatic Clams and Zebra zebra mussels, clams, fish, Mussels) has been identified as a potential hazard to the QCNPS algae) on an intake structure or RHR service water and diesel generator cooling water systems.
| |
| internal to a system that uses raw QCNPS has implemented a program to trend flow blockage cooling water from a source of characteristics. This information is used to ensure flow blockage will surface water, causing its not occur in safety-related systems using river water. The program Biological Events functional failure. C1 includes:
| |
| Periodic inspection of the intake bays.
| |
| Periodic flushing of infrequently used or stagnant lines in safety-related service water systems.
| |
| Annual water and substrate sampling.
| |
| Periodic testing, inspection and cleaning of safety-related heat exchangers.
| |
| E4-48
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| Periodic inspection of high-flow and low-flow service water piping for corrosion, erosion, silting and biofouling.
| |
| A long-term program for ultrasonic test examination of cooling water lines associated with safety-related heat exchangers.
| |
| Corrosion coupons installed in DGCW and RHRSW piping.
| |
| Based on this review, the biological events hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| Removal of material from a Coastal erosion is a slowly developing event and could be mitigated shoreline of a body of water (e.g., or adequately responded to (C5).
| |
| river, lake, ocean) due to surface processes (e.g., wave action, Also, per UFSAR Section 2.4.4 (Reference [37]), the entire intake tidal currents, wave currents, flume from the crib house to the river's edge is stripped out to the drainage, or winds and including natural rock which is approximately elevation 557 feet. The natural riverbed scouring) that results in C1 river bottom between the river's edge and the main river channel Coastal Erosion damage to the foundation of varies in elevation from elevation 557 feet to elevation 565 feet, thus SSCs at or in the plants C5 preventing a direct flow of water from the main channel to the crib analyzed area, causing functional house during a broken dam condition.
| |
| failure.
| |
| UFSAR Section 3.7.3.3.3 states that the retaining wall was analyzed for both an OBE and a DBE. The resulting stresses were below the allowables for both cases.
| |
| E4-49
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| UFSAR Section 3.7.3.3.4 states that the earth embankment was found to be capable of resisting the sliding effects during an SSE for all three cases considered. The cases were (1) circular sliding surface; (2) plane sliding surface; and (3) block sliding horizontally (C1).
| |
| Based on this review, the coastal erosion hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| A shortage of surface water Drought is a slowly developing hazard allowing time for orderly plant supplies due to a period of reductions, including shutdowns.
| |
| below-average precipitation in a given region, thereby depleting Based on this review, the drought hazard can be considered Drought the water supply needed for the C5 negligible.
| |
| various water-cooling functions at the facility. There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| An excess of water outside the Based on the review in Section 5 of this enclosure, the risk from the plant boundary that causes external flood hazard can be considered negligible.
| |
| functional failure to plant SSCs. C1 External Flood External flood causes include, There are no configuration-specific considerations for this hazard.
| |
| but may not be limited to, PS4 This hazard can be excluded from the TSTF-505 program flooding due to dam failure, high evaluation.
| |
| tide, hurricane (tropical cyclone),
| |
| E4-50
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17) ice cover, local intense precipitation, river diversion, river and stream overflow, seiche, storm surge, and tsunami.
| |
| Strong winds resulting in dynamic See Section 4 of this Enclosure.
| |
| loading or missile impacts on SCCs causing functional failure.
| |
| Hazards that could potentially result in high wind include the following:
| |
| hurricane - severe winds developed from a tropical depression resulting in Extreme Winds and missiles or dynamic loading on N/A Tornadoes SSCs. Secondary hazards resulting from a hurricane, include, but are not necessarily limited to tornado straight wind - a strong wind resulting in missiles or dynamic loading on SSCs that is not associated with either hurricanes or tornadoes E4-51
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17) tornado - a strong whirlwind that results in missiles or dynamic loading on SSCs Fog Low-lying water vapor in the form of a cloud or obscuring haze of atmospheric dust or smoke resulting in impeded visibility that could result in, for example, a transportation accident.
| |
| Low-lying water vapor in the form The principal effects of such events (such as freezing fog) would be of a cloud or obscuring haze of to cause a loss of off-site power, which is addressed in weather-atmospheric dust or smoke related LOOP scenarios in the FPIE PRA model for QCNPS.
| |
| resulting in impeded visibility that Fog could result in, for example, a C4 Based on this review, the fog hazard can be considered negligible.
| |
| transportation accident.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| Direct (e.g., thermal effects) and Per IPEEE Table 5-1 (Reference [7]), the site landscaping and lack indirect effects (e.g., generation of forestation prevent such fires from posing a threat to QCNPS of combustion products, transport (C3).
| |
| of firebrand) of a forest fire C3 Forest Fire outside the plant boundary that In addition, forest fires originating from outside the plant boundary causes functional failure of plant C4 may cause a loss of offsite power event, which is addressed for SSCs. grid-related LOOP scenarios in the FPIE PRA model for QCNPS (C4).
| |
| E4-52
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| Hazards that could cause or be Based on this review, the forest fire hazard can be considered caused by a forest fire include, negligible.
| |
| but may not be limited to, wildfires and grass fires. There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| A thin layer of ice crystals that The principal effects of such events would be to cause a loss of form on the ground or the surface off-site power, which is addressed for weather-related LOOP of an earthbound object when the scenarios in the FPIE PRA model for QCNPS.
| |
| temperature of the ground or Frost surface of the object falls below C4 Based on this review, the frost hazard can be considered negligible.
| |
| freezing.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| A shower of ice or hard snow that The principal effects of such events would be to cause a loss of could result in transportation off-site power, which is addressed for weather-related LOOP accidents or directly causes scenarios in the FPIE PRA model for QCNPS. (C4) dynamic loading or freezing conditions as a result of ice Flooding impacts are covered under external flooding/intense C1 coverage. precipitation. (C1)
| |
| Hail C4 Based on this review, the hail hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| E4-53
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| Effects on SSC operation due to Per UFSAR Section 1.2.1 (Reference [37]), the design of abnormally high ambient components important to safety of the units and the station includes temperatures resulting from allowances for environmental phenomena (C1).
| |
| weather phenomena. Secondary hazards resulting from high Per UFSAR Section 9.2.5.3 (Reference [37]), recirculation of the ambient temperatures, include, ultimate heat sink (UHS) volume would result in elevated RHRSW but are not necessarily limited, to and DGCW temperatures. The worst-case scenario would be a dam low lake or river water levels. failure during the summer months with a river temperature of 95°F and minimum evaporative cooling. Assuming a conservative ultimate heat sink volume of 2.15 million gallons of water, 5100 gpm of makeup from the portable pumps and 24 hours of shutdown time on the main condenser, this results in a maximum RHRSW and C1 DGCW inlet temperature of 109°F. This would be sufficient for each High Summer unit to operate one RHR pump, one RHR heat exchanger, one C4 Temperature RHRSW pump and one DGCW pump for a total flow of approximately 10,000 gpm for cooling (C1).
| |
| C5 In addition, continuous use procedure QCOP 0010-10, "Required Hot Weather Inspections" (Reference [59]) contains guidance for inspections of various plant locations and equipment during hot weather conditions (C5).
| |
| Plant trips due to this hazard are covered in the definition of another event in the PRA model (e.g., transients, loss of condenser) (C4).
| |
| Based on this review, the high summer temperature hazard can be considered negligible.
| |
| E4-54
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| The periodic maximum rise of Since the site is located on the Mississippi River, the high tide sea level resulting from the hazard does not apply (C3).
| |
| combined effects of the tidal gravitational forces exerted by Based on this review, the high tide hazard can be considered High Tide the Moon and Sun and the C3 negligible.
| |
| rotation of the Earth.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| Flooding that results from the Hurricanes are extreme tropical storms that originate offshore and intense rain fall from a hurricane as such do not reach QCNPS due to the mid-western location of the (tropical cyclone). site (C3).
| |
| Hurricane (Tropical Secondary hazards resulting Based on this review, the hurricane hazard can be considered C3 Cyclone) from a hurricane include, but are negligible.
| |
| not necessarily limited to, dam failure, high tide, river and stream There are no configuration-specific considerations for this hazard.
| |
| overflow, seiche, storm surge, This hazard can be excluded from the TSTF-505 program and waves. evaluation.
| |
| Flooding due to downstream UFSAR Section 2.4.7 (Reference [37]) discusses ice cover. An ice-blockages of ice on a river. C1 melting line is tied into the side of the discharge flume upstream of a Ice Cover weir with a top elevation at 574.75 feet. This line is a 96-inch Secondary hazards resulting C4 diameter pipe with a bottom elevation at that point of 557 feet. A from an ice blockage include, but gate is provided in this line for shutoff. The ice-melting line runs into E4-55
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17) are not necessarily limited to, and across the intake flume and is provided with four outlets having river and stream overflow. a bottom elevation of 558 feet. During frozen conditions, the circulating water system would be in service, and the ice melting line would normally be opened enough to keep the intake forebay free of ice (C1).
| |
| The principal effect of ice cover events would be to cause a loss of off-site power event, which is addressed for weather-related LOOP scenarios in the FPIE PRA model for QCNPS (C4).
| |
| Based on this review, the ice cover hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| An accident at an offsite Per the IPEEE (Reference [7]), and UFSAR Section 2.2 industrial or military facility that (Reference [37]), there are no military facilities within a 5-mile radius results in a release of toxic of QCNPS (C3).
| |
| gases, a release of combustion products, a release of UFSAR Section 2.2 further states that within a 5-mile radius of C1 Industrial or Military radioactivity, an explosion, or the QCNPS, the general character of land use is rural, comprised of Facility Accident generation of missiles. scattered villages and homes, except for two industrial areas.
| |
| C3 The industrial areas located within a 5-mile radius of include the Cordova Industrial Park, which has the nearest major industrial tenant, the Minnesota Mining and Manufacturing (3M) plant, a chemical company situated 1 1/2 miles from the site. There is also E4-56
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17) the CF Industries Chemical complex, located 3.1 miles north of the site.
| |
| Per UFSAR Section 2.2.2, none of the operations at Cordova Industrial Park pose any threat to QCNPS from explosion, explosive shock, resulting missiles, or toxic fumes release. Furthermore, there is no chlorine gas used in the area (C1, C3).
| |
| In addition, to mitigate the impact of an industrial facility accident, QCOA 0010-17, "Toxic Gas/Chemical Release from Nearby Facilities" (Reference [60]) includes automatic (control room ventilation system isolation) and subsequent operator actions (don air packs, obtain temporary ventilation, etc.).
| |
| See also Toxic Gas.
| |
| Based on this review, the industrial or military facility accident hazard can be considered negligible.
| |
| There are no configuration specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| The QCNPS Internal Events PRA includes evaluation of risk from Internal Flood N/A internal flooding events.
| |
| The QCNPS Internal Fire PRA model addresses risk from internal Internal Fire N/A fires E4-57
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| Dynamic loading of SSCs or Plant site is located on level terrain and is not subject to landslides.
| |
| impacts on natural water supplies Additionally, the mid-western location of QCNPS precludes the used for heat rejection due to the possibility of a landslide.
| |
| movement of rock, soil, and mud down a sloped surface (does not Based on this review, the landslide hazard can be considered Landslide C3 include frozen precipitation). negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| Effects on SSCs due to a sudden Lightning strikes are not uncommon in nuclear plant experience.
| |
| electrical discharge from a cloud They can result in losses of off-site power or surges in to the ground or Earth-bound instrumentation output if grounding is not fully effective. The latter object. events often lead to reactor trips. Both events are incorporated into the QCNPS internal events model through the incorporation of generic and plant-specific data.
| |
| Lightning C4 Based on this review, the lightning hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| A decrease in the water level of Per UFSAR Section 9.2.5.2 (Reference [37]), the UHS is provided to the lake or river used for power C1 mitigate the consequences of the postulated failure of Lock and Dam Low Lake or River generation. No. 14 on the Mississippi River downstream of the plant, which Water Level C5 would cause river level to drop. The station design is such that if Lock and Dam No. 14 were to fail, the water level would recede in E4-58
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17) the intake bay to the point where it would be separated from the river.
| |
| UFSAR Section 9.2.5.2 states that the ultimate heat sink is defined as the water that is captured inside the log boom located in the discharge canal. This captured volume is approximately 3 million gallons (C1). Use of the ultimate heat sink to shut down the reactors requires the operation of portable diesel pumps along with RHRSW and DGCW to shutdown the units. Following a dam failure, approximately 2 days would be available to position the portable pumps (C5).
| |
| Per UFSAR Section 2.4.4, if a natural disaster such as an earthquake were to occur, the dam would most likely not incur the complete loss of a function. Such a disaster would likely result in malfunctioning or loss of gate operating capability. This condition would not result in the loss of the pool used as the plants ultimate heat sink but simply the inability to operate the Lock & Dam gates.
| |
| Therefore, the maximum credible failure is considered the loss of both the upstream and downstream lock miter gates (C1).
| |
| Continuous use Procedure QCOA 0010-14, "Lock and Dam #14 Failure" (Reference [61]) also provides direction for loss of river due to dam failure. In the procedure, use of the ultimate heat sink requires taking in River water through the discharge piping and returning it over the log boom at the intake flume.
| |
| Based on this review, the Low Lake or River Water Level hazard can be considered negligible.
| |
| E4-59
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| Effects on SSC operation due to Per UFSAR Section 1.2.1 (Reference [37]), the design of abnormally low ambient components important to safety of the units and the station includes temperatures resulting from allowances for environmental phenomena (C1).
| |
| weather phenomena.
| |
| In addition, continuous use procedure QCOP 0010-02, "Required Secondary hazards resulting Cold Weather Inspections" (Reference [62]) contains guidance for from low ambient temperatures inspections of various plant locations and equipment during cold include, but are not necessarily weather conditions (C5).
| |
| limited to, frost, ice cover, and C1 snow. In addition, plant trips due to this hazard are covered in the definition Low Winter C4 of another event in the PRA model (e.g., transients, loss of Temperature condenser) (C4).
| |
| C5 See also Ice Cover.
| |
| Based on this review, the low winter temperature hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| E4-60
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| A release of energy due to the Per the IPEEE (Reference [7]), the frequency of a meteor or satellite impact of a space object such as strike is judged to be so low as to make the risk impact from such a meteoroid, comet, or human- events insignificant.
| |
| caused satellite falling within the Earths atmosphere, a direct Based on this review, the meteorite or satellite hazard can be impact with the Earths surface, considered negligible.
| |
| or a combination of these effects.
| |
| Meteorite/Satellite There are no configuration-specific considerations for this hazard.
| |
| PS4 Strikes This hazard is analyzed with This hazard can be excluded from the TSTF-505 program respect to direct impacts of an evaluation.
| |
| SSC and indirect impact effects such as thermal effects (e.g.,
| |
| radiative heat transfer),
| |
| overpressure effects, seismic effects, and the effects of ejecta resulting from a ground strike.
| |
| E4-61
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| A release of hazardous material, Per IPEEE Table 5-1(Reference [7]), the nearest pipelines are one a release of combustion mile from the plant and consist of a 3-inch diameter and a 12-inch products, an explosion, or the diameter pipe. Both are natural gas pipelines. There is no generation of missiles due to an significant hazard to the plant from these lines. Other lines, located accident involving the rupture of three miles or more from the site, carry ethane, ethylene, and a pipeline carrying hazardous propane through 12-inch pipes. These lines have been evaluated in materials. a "Control Room Habitability Study for QCNPS Units 1 and 2 as cited in Section 2.2 of the UFSAR (Reference [37]). No significant toxic hazard exists from pipelines near the site (C1, C3).
| |
| C1 See also Industrial or Military Facility Accident, Toxic Gas, and Pipeline Accident Transportation Accidents.
| |
| C3 Based on this review, the pipeline accident hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| E4-62
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| Flooding that results from local See Section 5 of this report.
| |
| intense precipitation.
| |
| The IA (Reference [45]) confirmed that the site has provided Secondary hazards resulting adequate protection from the LIP flooding mechanism. There are no from local intense precipitation, impacts to the plant once the six LIP barriers are installed (Table E4-include, but are not necessarily 12: QCNPS LIP Flood Barriers).
| |
| Precipitation, limited to, dam failure and river C1 Intense and stream overflow. Based on this review, the risk from the intense precipitation hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| A release of hazardous material Per IPEEE Table 5-1(Reference [7]), chlorination of water systems is including, but not limited to performed using a hypochlorite system. No chlorine gas is stored liquids, combustion products, or onsite. Various acids and caustics are stored on-site but pose no radioactivity. hazard to the plant. Hydrogen and liquid oxygen are stored on-site but compliance with EPRI guidelines ensures that no significant Such releases may be concurrent hazard results from these materials.
| |
| Release of with or induce an explosion or the Chemicals from generation of missiles. C1 Per USFAR Sections 6.4.4.2.1- 6.4.4.2.2 (Reference [37]), models Onsite Storage were developed to calculate the concentrations of toxic chemicals in In this context, an onsite release the control room in the event of an accidental spill consistent with of radioactivity is assumed to be the models described in NUREG-0570 (Reference [63]).
| |
| associated with low-level radioactive waste. Based on the physical and toxicological properties of the chemicals stored at the QCNPS site, it is concluded that none of the chemicals are of concern. For these chemicals, the unisolated control room E4-63
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17) concentrations will not exceed the threshold limit value in the event of a postulated release.
| |
| See also Toxic Gas.
| |
| Based on this review, the release of chemicals from onsite storage hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| The redirection of all or a portion Per UFSAR 2.4.9 (Reference [37]), the authority to control the river of river flow by natural causes is vested in the U.S. Army Corps of Engineers. Should the need to (e.g., a riverine embankment control the river arise, Constellation will make the required landslide) or intentionally (e.g., notification to the Corps of Engineers. These arrangements have power production, irrigation). been made and are detailed in the QCNPS emergency procedures.
| |
| The NRC confirmed in its evaluation of the QCNPS Flood Hazard Reevaluation Report that flooding from channel migrations or River Diversion C1 diversions is not a plausible flood hazard mechanism at the QCNPS site (Reference [56]).
| |
| Based on this review, the river diversion hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| E4-64
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| Persistent heavy winds Per the IPEEE (Reference [7]), the midwestern location of QCNPS transporting sand or dust that prevents sandstorms. More common wind-borne dirt can occur but infiltrate SSCs at or in the plants poses no significant risk given the robust structures and protective analyzed area causing functional features of the plant.
| |
| failure.
| |
| Sandstorm C1 Based on this review, the sandstorm hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| Flooding from water displaced by Per UFSAR Section 2.4.5 (Reference [37]), flooding due to seiches an oscillation of the surface of a is not applicable to QCNPS.
| |
| landlocked body of water, such as a lake, that can vary in period In its review of the QCNPS Flood Hazard Reevaluation Report from minutes to several hours. (Reference [56]), the NRC confirmed that flooding due to seiche is not a plausible flooding mechanism due to the location of the site in relation to large bodies of water, and therefore does not impact the Seiche C3 site.
| |
| Based on this review, the seiche hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| E4-65
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| Sudden ground motion or See Section 3 of this Enclosure.
| |
| vibration of the Earth as produced by a rapid release of stored-up energy along an active fault.
| |
| Secondary hazards resulting Seismic Activity N/A from seismic activity include, but are not necessarily limited to, avalanche (both rock and snow),
| |
| dam failure, industrial accidents, landslide, seiche, tsunami, and vehicle accidents.
| |
| The accumulation of snow could This hazard is slow to develop and can be identified via monitoring result in transportation accidents and managed via normal plant processes (C5).
| |
| or directly cause dynamic loading or freezing conditions as a result Potential flooding impacts are accounted for under external flooding of snow cover. C4 screening (C4).
| |
| Snow C5 Based on this review, the snow hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| E4-66
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| Dynamic forces on structures The potential for this hazard is low at the site, the plant design foundations due to the expansion considers this hazard (C1), and the hazard is slow to develop and (swelling) and contraction can be mitigated (C5).
| |
| (shrinking) of soil resulting from C1 changes in the soil moisture Based on this review, the soil shrink-swell consolidation impact Soil Shrink-Swell content. hazard can be considered negligible.
| |
| C5 There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| Flooding that results from an Per UFSAR Section 2.4.5 (Reference [37]), flooding due to surges is abnormal rise in sea level due to not applicable to QCNPS.
| |
| atmospheric pressure changes and strong wind generally In its review of the QCNPS Flood Hazard Reevaluation Report accompanied by an intense (Reference [56]), the NRC confirmed the location of the site in storm. relation to large nearby waterbodies and the site is approximately 140 miles from the coast of Lake Michigan, which would be only Secondary hazards resulting source of a storm surge, and the topography between the site and Storm Surge C3 from a storm surge include, but Lake Michigan would attenuate a storm surge and the associated are not necessarily limited to, wave runup (Reference [56]). The NRC confirmed the licensee's high tide, river and stream conclusion that storm surge is not a plausible flood hazard overflow, and waves. mechanism at QCNPS and would not impact the site.
| |
| Based on this review, the storm surge hazard can be considered negligible.
| |
| E4-67
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| A release of hazardous toxic or Per USFAR Section 2.2.3.2 (Reference [37]), liquid hydrogen and asphyxiant gases. liquid oxygen storage facilities are installed at the site. Compliance with EPRI guidelines (Reference [64]) ensures that the system Such releases may be concurrent installation and operation will not produce a safety concern.
| |
| with or induce an explosion or the Additionally, the onsite delivery routes for transporting hydrogen and generation of missiles. oxygen to their respective storage facilities comply with EPRI guidelines and therefore will not produce a safety concern.
| |
| In this context, an onsite release of radioactivity is assumed to be USFAR Section 6.4.4.2 (Reference [37]) outlines the toxic gas associated with low-level analysis. Per USFAR Section 6.4.4.2.2, based on the physical and radioactive waste. toxicological properties of the chemicals stored at the QCNPS site, it is concluded that none of the chemicals are of concern. For these Toxic Gas C1 chemicals, the unisolated control room concentrations will not exceed the threshold limit value (TLV) in the event of a postulated release.
| |
| Per USFAR Section 6.4.4.2.3, the control room HVAC system provides toxic gas protection to the control room emergency zone in case of either an onsite or offsite toxic chemical accident. The system provides this protection by either manual isolation through operator action or automatic isolation via a toxic gas analyzer.
| |
| For ammonia, a monitor is provided since the control room concentrations reach toxicity limits faster than the operator can manually isolate the system after detection of odor.
| |
| E4-68
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| See also Transportation Accidents and Release of Chemicals from Onsite Storage.
| |
| Based on this review, the toxic gas hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| Accidents involving transportation USFAR Section 2.2 (Reference [37]) states that there are six resulting in collision with SSCs, a transportation routes within a 5-mile radius of the plant: the release of hazardous materials or Mississippi River, U.S. Route 67, State Route 84, and three railroad combustion products, an lines.
| |
| explosion, or a generation of missiles causing functional failure Per the IPEEE (Reference [7]), rail, barge, aircraft, and pipeline of SSCs. transportation accidents are insignificant hazards to the site. (C1, C1 C3)
| |
| Transportation Hazards that could potentially C3 Accidents result in transportation accidents In addition, per USFAR Section 6.4.4.2.3, the control room HVAC include, for example, a vehicle, system provides toxic gas protection to the control room emergency C4 railcar or ship (boat) accident that zone in case of either an onsite or offsite toxic chemical accident.
| |
| involves a collision or derailment, The system provides this protection by either manual isolation potentially resulting in fire, through operator action or automatic isolation via a toxic gas explosions, toxic releases, analyzer (C1).
| |
| missiles, or other hazardous conditions. Other releases of toxic gases during transportation to local facilities are included in the toxic gas hazard screening (C1, C4).
| |
| E4-69
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| See also Toxic Gas.
| |
| Based on this review, the transportation accidents hazard can be considered negligible.
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| Flooding that results from a Per UFSAR Section 2.4.6 (Reference [37]), flooding due to tsunamis series of long-period sea waves is not applicable to QCNPS.
| |
| that displaces massive amounts of water as a result of an In its review of the QCNPS Flood Hazard Reevaluation Report impulsive disturbance, such as a (Reference [56]), the NRC confirmed flooding due to tsunamis is not major submarine slides or applicable to QCNPS.
| |
| Tsunami landslide. C3 Based on this review, the tsunami hazard can be considered Secondary hazards resulting negligible.
| |
| from a tsunami include, but are not necessarily limited to, river There are no configuration-specific considerations for this hazard.
| |
| and stream overflow. This hazard can be excluded from the TSTF-505 program evaluation.
| |
| Damage to safety-related SSCs For main turbine missiles, the NRC-preferred option for unfavorably from a missile generated internal oriented main turbines such as QCNPS (i.e., main turbine orientation Turbine-Generated or external to the plant PRA such that essential SSCs are inside the RG 1.115 (Reference [65])
| |
| PS3 Missiles boundary from rotating turbines defined turbine missile strike zone), is to limit P1, the annual or other external sources (e.g., probability of "low trajectory" turbine missile generation resulting in high-pressure gas cylinders).
| |
| E4-70
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| Damage may result from a falling the ejection of turbine disk (or internal structure) fragments through missile or a missile ejected the turbine casing, to 1E-5.
| |
| directly toward safety-related SSCs (i.e., low-trajectory A 2019 evaluation (performed consistent with RG 1.115 approaches) missiles). was performed to calculate the P1 value for various cases of test intervals of turbine overspeed protection (OPS) components for the QCNPS main turbine (Reference [66]).
| |
| The value of P1 for QCNPS is the sum of the annual probabilities of the two RG 1.115 missile generation scenarios: 1) brittle fracture probability at design speed and 2) ductile failure probability given turbine overspeed which increases when main turbine overspeed protection system components are tested less frequently. Based on the monthly main turbine surveillance tests corresponding to the "Case 3 Test Interval" from Table 6-3 of Reference [66], the total P1 probability is 8.01E-6.
| |
| Per Table 1 of RG 1.115, the likelihood of main turbine missile impact on essential equipment is 1E-7/yr for a plant with an unfavorably oriented main turbine demonstrating a P1 value of less than 1E-5/yr. As such, the estimated CDF from a QCNPS postulated main turbine missile is less than 1E-7/yr.
| |
| Based on this review, the turbine missile hazard can be considered negligible.
| |
| There are no configuration specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| E4-71
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| Opening of Earths crust resulting This hazard is not applicable to the site because of location (no in tephra (i.e., rock fragments active or dormant volcanoes located near the plant site).
| |
| and particles ejected by volcanic eruption), lava flows, lahars (i.e., Based on this review, the volcanic activity hazard can be considered mud flows down volcano slopes), negligible.
| |
| volcanic gases, pyroclastic flows (i.e., fast-moving flow of hot gas There are no configuration-specific considerations for this hazard.
| |
| and volcanic matter moving down This hazard can be excluded from the TSTF-505 program and away from a volcano), and evaluation.
| |
| landslides.
| |
| Volcanic Activity C3 Indirect impacts include distant ash fallout (e.g., tens to potentially thousands of miles away).
| |
| Secondary hazards resulting from volcanic activity, include, but are not necessarily limited to, seismic activity and fire.
| |
| An area of moving water that is Refer to Section 5 of this Enclosure. Wind-generated waves are raised above the main surface of subsumed in the combined effects flood mechanism which was a body of water as a result of the screened with a CDF well below 1E-6/yr.
| |
| Waves wind blowing over an area of fluid PS4 surface. Based on this review, the waves hazard can be considered negligible.
| |
| E4-72
| |
| | |
| License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Enclosure 4 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-16: Other External Hazards Disposition3 Screening Hazard Definition Criteria QCNPS Disposition for TSTF-505 (Table E4-17)
| |
| There are no configuration-specific considerations for this hazard.
| |
| This hazard can be excluded from the TSTF-505 program evaluation.
| |
| E4-73
| |
| | |
| License Amendment Request Enclosure 4 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-17: Progressive Screening Approach for Addressing External Hazards Event Analysis Criterion Source C1. Event damage potential is < NUREG/CR-2300 and ASME/ANS events for which plant is designed. Standard RA-Sa-2009 C2. Event has lower mean frequency and no worse NUREG/CR-2300 and ASME/ANS consequences than other events Standard RA-Sa-2009 Initial analyzed.
| |
| Preliminary C3. Event cannot occur close NUREG/CR-2300 and ASME/ANS Screening enough to the plant to affect it. Standard RA-Sa-2009 C4. Event is included in the NUREG/CR-2300 and ASME/ANS definition of another event. Standard RA-Sa-2009 C5. Event develops slowly, allowing adequate time to eliminate ASME/ANS Standard RA-Sa-2009 or mitigate the threat.
| |
| Progressive PS1. Design basis hazard cannot ASME/ANS Standard RA-Sa-2009 Screening cause a core damage accident.
| |
| PS2. Design basis for the event NUREG-1407 and ASME/ANS meets the criteria in the NRC 1975 Standard RA-Sa-2009 Standard Review Plan (SRP).
| |
| PS3. Design basis event mean frequency is < 1E-5/y and the mean NUREG-1407 as modified in conditional core damage probability ASME/ANS Standard RA-Sa-2009 is < 0.1.
| |
| PS4. Bounding mean CDF is NUREG-1407 and ASME/ANS
| |
| < 1E-6/y. Standard RA-Sa-2009 Screening not successful. PRA NUREG-1407 and ASME/ANS Detailed PRA needs to meet requirements in the Standard RA-Sa-2009 ASME/ANS PRA Standard.
| |
| 7 Conclusions Based on this analysis of external hazards for QCNPS, no additional external hazards other than seismic and extreme wind events for tornado missiles need to be added to the existing PRA model. The evaluation concluded that the hazards either do not present a design-basis challenge to QCNPS, the challenge is adequately addressed in the PRA, or the hazard has a negligible impact on the calculated RICT and can be excluded.
| |
| Therefore, QCNPS will apply seismic and high wind penalties in the risk evaluations performed as part of the process to calculate a RICT. In this application, all other external hazards are considered insignificant and will not be included in the RICT calculation.
| |
| The ICDP/ILERP acceptance criteria of 1E-5/1E-6 will be used within the PARAGON framework to calculate the resulting RICT and RMAT based on the total configuration-specific delta E4-74
| |
| | |
| License Amendment Request Enclosure 4 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models CDF/LERF attributed to internal events and internal fire, plus the seismic and tornado risk bounding delta CDF/LERF values.
| |
| 8 References
| |
| [1] Nuclear Energy Institute (NEI) 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," Revision 0-A, dated October 12, 2012 (ADAMS Accession No. ML12286A322)
| |
| [2] Letter from J.M. Golder (NRC) to B. Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines',"
| |
| dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| [3] ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum A to RAS-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009
| |
| [4] NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," Revision 1, dated March 2017 (ADAMS Accession No. ML17062A466)
| |
| [5] NUREG-75/087, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," dated 1975 (ADAMS Accession Nos. ML042080427, ML0422080217, ML042080430, and ML042080090) [Note this document has been reissued as NUREG-0800]
| |
| [6] NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," dated June 1991 (ADAMS Accession No. ML063550238)
| |
| [7] Letter from R.M Krich (Exelon Generation Company, LLC) to U.S. NRC, "Updated Individual Plant Examination of External Events Report," dated July 29, 1999
| |
| [8] Exelon Generation Company, LLC Report No. SL-012196, "Seismic Hazard and Screening Report - Quad Cities Units 1 and 2," dated March 2014
| |
| [9] Information Notice (IN) 2010-18, "Generic Issue (GI) 199 - Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants," Tables B.2, C.1 and C-2, dated September 2, 2010, (ADAMS Accession No. ML100270582)
| |
| E4-75
| |
| | |
| License Amendment Request Enclosure 4 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
| |
| [10] Letter RS-14-072 from G. Kaegi (Exelon Generation Company, LLC) to U.S. NRC, "Seismic Hazard and Screening Report (Central and Eastern United States (CEUS) Sites),
| |
| "'Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident'," dated March 31, 2014 (ADAMS Accession No. ML14090A526)
| |
| [11] EPRI 3002000709, "Seismic Probabilistic Risk Assessment Implementation Guide,"
| |
| Electric Power Research Institute, dated December 2013
| |
| [12] Letter from U.S. NRC to O.D. Kingsley (Exelon Generation Company, LLC), "Quad Cities Nuclear Power Plant - Review of Individual Plant Examination of External Events (IPEEE)
| |
| Submittal (Tac Nos. M83665 and M83666)," dated April 26, 2001
| |
| [13] Letter RS-12-169 from G. Kaegi (Exelon Generation Company, LLC) to U.S. NRC, "Exelon Generation Company, LLC's 180-day Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated November 27, 2012 (ADAMS Accession No. ML12361A097)
| |
| [14] Electric Power Research Institute (EPRI) NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision 1, dated August 1991 (publically available at https://www.epri.com/research/products/NP-6041-SLR1)
| |
| [15] NUREG/CR-0098, "Development of Criteria for Seismic Review of Selected Nuclear Power Plants," dated May 1978 (ADAMS Accession No. ML061880412)
| |
| [16] Quad Cities Full-Power Internal Events PRA Model of Record (MOR), QC118A
| |
| [17] Letter NL-12-1344 from M.J. Ajluni (Southern Company) to U.S. NRC, "Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, Risk-Informed Technical Specifications Initiative 4b Risk-Managed Technical Specifications (RMTS) Guidelines,"
| |
| dated September 13, 2012 (NRC ADAMS Accession No. ML12258A055)
| |
| [18] Letter from D.P. Helker (Exeon Generation Company, LLC) to U.S. NRC, "License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 1, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b.'," dated February 25, 2016 (ADAMS Accession No. ML16060A223)
| |
| [19] EPRI 1025287, "Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," Electric Power Research Institute, dated February 2013 (ADAMS Accession No. ML12333A170)
| |
| E4-76
| |
| | |
| License Amendment Request Enclosure 4 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
| |
| [20] Enclosure B, "Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1, Beaver Valley Power Station Unit No. 2," dated July 2017 (ADAMS Accession No. ML17213A017)
| |
| [21] Letter CNL-19-122 from J.T. Polickoski (TVA) to U.S. NRC, "Seismic Probabilistic Risk Assessment for Browns Ferry Nuclear Plant, Units 1, 2 and 3 - Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommentation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident," dated December 2019 (ADAMS Accession No. ML19351E391)
| |
| [22] Attachment 2 to ULNRC-06524, "Callaway Energy Center Seismic Probabilistic Risk Assessment in Response to 10 CFR 50.54(f) Letter with Regard to NTTF 2.1: Seismic,"
| |
| dated August 2019, (ADAMS Accession No. ML19225D324)
| |
| [23] Letter GO2-19-136 from J.K. Ditmer (Energy Northwest) to U.S. NRC, "Columbia Generating Station Docket No. 50-397 Seismic Probabilistic Risk Assessment Response to NRC RFI Pursuant to 10 CFR 50.54(f) Regarding Recommentation 2.1 of the NTTF Review of Insights from the Fukushim Dai-ichi Accident (MF3726, MF3727)," dated September 2019 (ADAMS Accession No. ML19273A907)
| |
| [24] Letter AEP-NRC-2019-58 from Q.S. Lies (AEP) to U.S. NRC, "DC Cook Nuclear Plant Units 1and 2 - Seismic Probabilistic Risk Assessment in Response to NTTF Recommendation 2.1: Seismic," dated November 2019 (ADAMS Accession No. ML19310D805)
| |
| [25] Letter DCL-15-035 from B.S. Allen (Pacific Gas & Electric Company) to U.S. NRC, "Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Seismic Aspects of Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dia-ichi Accident: Seismic Hazard and Screening Report," dated March 2015 (ADAMS Accession No. ML15070A607)
| |
| [26] Letter DCL-18-027 from J.M. Welsch (Pacific Gas and Electric Company) to U.S. NRC, "Seismic Probabilistic Risk Assessment for the Diablo Canyon Power Plant, Units 1 & 2 -
| |
| Response to NRC RFI Pursuant to 10 CFR 50.54(f) Regarding Recommenation 2.1:
| |
| Seismic of the NTTF Review of Insights from the Fukushima Dai-ichi Accident," dated April 24, 2018 (ADAMS Accession No. ML18120A201)
| |
| [27] DR-PRA-020.005, Volume 1, , "Dresden Seismic Probabilistic Risk Assessment Fragility Modeling Notebook,," Revision 1, dated October 2019 E4-77
| |
| | |
| License Amendment Request Enclosure 4 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
| |
| [28] Letter 18-086 from D.G. Stoddard (Virgina Electric Power Company) to U.S. NRC, "North Anna Power Station Units 1 and 2 Response to March 12, 2012 Information Request Seismic Probabilistic Risk Assessment for Recomendation 2.1," dated March 2018 (ADAMS Accession No. ML18093A445)
| |
| [29] PB-PRA-20.005, Volume 1, "Peach Bottom Seismic Probabilistic Risk Assessment Fragility Modeling Notebook," Revison 2, dated August 2018
| |
| [30] Letter CNL-19-061 from J.T. Polickoski (TVA) to U.S. NRC, "Seismic Probabilistic Risk Assessment for Sequoyah Nuclear Plant, Units 1 & 2 - Response to NRC RFI Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident," dated October 18, 2019 (ADAMS Accession No. ML19291A003)
| |
| [31] Letter RC-18-0117 from G.A. Lippard (SCE&G) to U.S. NRC, "Virgil C. Summer Nuclear Station, Unit 1, Docket No. 50-395, Operating License No. NPF-12, Fukushima NTTF Recommentation 2.1: Seismic, Seismic Probablistic Risk Assessment," dated September 28, 2018 (ADAMS Accession No. ML18271A109)
| |
| [32] Letter NL-17-0218 from J.J. Hutto (Souther Nuclear) to U.S. NRC, "Vogtle Electric Generating Plant Units 1 and 2 - Fukushima NTTF Reomendation 2.1: Seismic, Seismic Probabilistic Risk Assessment," dated March 27, 2017 (ADAMS Accession No. ML17088A130)
| |
| [33] Letter CNL-17-071 from J.W. Shea (TVA) to U.S. NRC, "Seismic Probabilistic Risk Assessment for Watts Bar Nuclear Plant, Units 1 and 2 - Response to NRC RFI Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident," dated June 30, 2017 (ADAMS Accession No. ML17181A485)
| |
| [34] E-mail from R. Drsek (FENOC) to V. Andersen (Jensen Hughes, Exelon RM), RE: BV failure to scram Am>2g?, 2/7/2020 [e-mail clarifies that failure to scram fragility used in Beaver Valley SPRA NTTF 2.1 submittal was set to the fragility screening level and equals Am=1.52g, Br=0.24 and Bu=0.32]
| |
| [35] NEI 91-04, "Severe Accident Issue Closure Guidelines, Nuclear Energy Institute,"
| |
| Revision 1, dated December 1994 (ADAMS Accession No. ML072850981)
| |
| [36] U.S. NRC, "Risk Assessment of Operational Events, Volume 2 - External Events -
| |
| Internal Fires - Internal Flooding - Seismic - Other External Events - Frequencies of Seismically-Induced LOOP Events (RASP Handbook)," Revision 1.02, dated November 2017 (ADAMS Accession No. ML17349A301)
| |
| E4-78
| |
| | |
| License Amendment Request Enclosure 4 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
| |
| [37] Quad Cities Nuclear Power Station , "Updated Final Safety Analysis Report," Revison 16, dated October 2021
| |
| [38] NUREG/CR-4461, "Tornado Climatology of the Contiguous United States," Revision 2, dated February 2007 (ADAMS Accession No. ML070810400)
| |
| [39] ASCE 7 Hazard Tool, in https://asce7hazardtool.online/
| |
| [40] QC-MISC-048, "Quad Cities Nuclear Power Station Tornado Missile Risk Assessment,"
| |
| Revision 0, dated May 2023
| |
| [41] NEI 17-02 , "Tornado Missile Risk Evaluator (TMRE) Industry Guidance Document,"
| |
| Revision 1, dated September 2017 (ADAMS Accession No. ML17268A036)
| |
| [42] Letter RS-13-047 from G. Kaegi (Exelon Generation Company, LCC) to U.S. NRC, "Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 2, Flooding Hazard Reevaluation Report," dated March 12, 2013 (ADAMS Accession No ML13081A037)
| |
| [43] Letter from T. Govan (U.S. NRC) to B.C. Hanson (Exelon Generation Company, LLC),
| |
| "Quad Cities Nuclear Power Station, Units 1 ad 2 - Staff Assessment of Response to 10 CFR 50.54(f) Information Request - Flood Causing Mechanism Reevaluation (CAC Nos. MF1108 and MF1109)," dated November 18, 2016 (ADAMS Accession No. ML16323A343)
| |
| [44] Letter RS-18-045 from D.L. Helker (Exelon Generation Company, LLC) to U.S. NRC, "Exelon Generation Company, LLC Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Integrated Assessment Submittal," dated June 29, 2018 (ADAMS Accession No. ML18180A033)
| |
| [45] NEI 16-05, "External Flooding Assessment Guidelines," Revision 1, dated June 2016 (ADAMS Accession No. ML16165A178)
| |
| [46] Letter from M.J. Ross-Lee (U.S. NRC) to B.C. Hanson (Exelon Generation Company, LLC), "Quad Cities Nuclear Power Station, Units 1 and 2 - Staff Assessment of Flood Hazard Integrated Assessment (EPID No. L-2018-JLD-0008)," dated August 29, 2019 (ADAMS Accession No. ML19168A196)
| |
| [47] QDC-0000-S-2089, "Evaluation of Flood Boundary Structures for Local Intense Precipitation," Revision 2, dated January 29, 2018
| |
| [48] QCOA 0010-22, "Local Intense Precipitation Response Procedure," Revision 11 E4-79
| |
| | |
| License Amendment Request Enclosure 4 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
| |
| [49] EC 404409, "Integrated Review of Flex Actions IAW NEI Validation Process Fukushima,"
| |
| dated April 20, 2016
| |
| [50] LN-FLEX.2, Licensed/Non-Licensed Operator Initial/Continual Training, FLEX Equipment, August 2019
| |
| [51] QCOA 0010-16, "Flood Emergency Procedure," Revision 29
| |
| [52] Quad Cities Nuclear Power Station, "Probabilistic Flood Hazard Assessment Report for the Mississippi River," dated December 7, 2021
| |
| [53] EC 636912, "Update to Station External Flood Response in Support of Risk," dated June 10, 2022
| |
| [54] EC 636914, "Update to LIP Barriers to Assist the Station External Flood," dated June 3, 2022
| |
| [55] EC 404409, "Integrated Review of FLEX Actions IAW NEI Validation Process Fukushima,"
| |
| dated February 1, 2016
| |
| [56] NRC Letter To Exelon Generation Company, LLC, , "Quad Cities Nuclear Power Station, Units 1 and 2 - Staff Assessment of Response to 10 CFR 50.54(f) Information Request -
| |
| Flood-Causing Mechanism Reevaluation (CAC Nos. MF1108 and MF1109)," dated November 18, 2016 (ADAMS Accession No. ML16323A343)
| |
| [57] ATADS data query, in https://adip.faa.gov/agis/public/#/airportData/CWI, (accessed 24 August 2022)
| |
| [58] ATADS data query, in https://adip.faa.gov/agis/public/#/airportData/MLI, (accessed 24 August 2022)
| |
| [59] QCOP 0010-10, "Required Hot Weather Inspections," Revision 26
| |
| [60] QCOA 0010-17, "Toxic Gas/Chemical Release From Nearby Facilities," Revision 9
| |
| [61] QCOA 0010-14, "Lock and Dam #14 Failure," Revision 12
| |
| [62] QCOP 0010-02, Required Cold Weather Inspections, Revision 63 E4-80
| |
| | |
| License Amendment Request Enclosure 4 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
| |
| [63] NUREG-0570, "Toxic Vapor Concentrations in the Control Room Following a Postulated Accidental Release," dated June 1979 (ADAMS Accession No. ML063480551)
| |
| [64] "Guidelines for Permanent BWR Hydrogen Water Chemistry Installations," dated December 1985
| |
| [65] Regulatory Guide 1.115, "Protection Against Turbine Missiles," Revision 2, dated January 2012 (ADAMS Accession No. ML101650675)
| |
| [66] QDC-5650-I-1515, "Missile Probability for Nuclear BWR Retrofit," Revision 0, dated November 2019
| |
| [67] NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," June 1991
| |
| [68] Quad Cities Nuclear Power Station, "Updated Final Safety Analysis Report," Revision 16, dated October 2021
| |
| [69] QC-PSA-015, "Quad Cities Probabilistic Risk Assessment Level 2 Notebook," Revision 8, dated June 2019
| |
| [70] QCNPS letter to NRC, "Exelon Generation Company, LLC Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Integrated Assessment Submittal," dated June 29, 2018 (ADAMS Accession No. ML18180A033)
| |
| [71] Regulatory Guide 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, dated December 2020 (ADAMS Accession No. ML20238B871)
| |
| E4-81
| |
| | |
| ENCLOSURE 5 License Amendment Request Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF)
| |
| | |
| License Amendment Request Enclosure 5 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF)
| |
| : 1. Introduction Section 4.0, Item 6 of the U.S. Nuclear Regulatory Commission (NRC) Final Safety Evaluation (Reference [1]) for NEI 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines" (Reference [2]), requires that the license amendment request (LAR) provide the plant-specific total core damage frequency (CDF) and total large early release frequency (LERF) to confirm applicability of the limits of Regulatory Guide (RG) 1.174, Revision 1 (Reference [3]). (Note that RG 1.174, Revision 3 (Reference [4]), issued by the NRC in January 2018, did not revise these limits.)
| |
| The purpose of this enclosure is to demonstrate that the Quad Cities Nuclear Power Station (QCNPS) total CDF and total LERF are below the guidelines established in RG 1.174.
| |
| RG 1.174 does not establish firm limits for total CDF and LERF, but it recommends that risk-informed applications be implemented only when the total plant risk is no more than about 1E-4/year for CDF and 1E-5/year for LERF. Demonstrating that these limits are met confirms that the risk metrics of NEI 06-09 can be applied to the QCNPS Risk Informed Completion Time (RICT) Program.
| |
| : 2. Technical Approach Table E5-1 lists the CDF and LERF point estimate values that resulted from a quantification of the baseline internal events (including internal flooding) and fire Probabilistic Risk Assessment (PRA) models (References [5] and [6]). This table also includes an estimate of the seismic contribution to CDF and LERF based on the methodology detailed in Enclosure 4, Section 3 (Reference [7]). Other external hazards are below accepted screening criteria and therefore do not contribute significantly to the totals.
| |
| Table E5-1: Total Baseline CDF/LERF QCNPS Unit 1 Baseline CDF QCNPS Unit 1 Baseline LERF Source Contribution Source Contribution Internal Events PRA 3.9E-06 Internal Events PRA 2.4E-07 Fire PRA 3.8E-05 Fire PRA 3.2E-06 Seismic 4.3E-06 Seismic 2.0E-06 No significant No significant Other External Events Other External Events contribution contribution Total CDF 4.7E-05 Total LERF 5.4E-06 QCNPS Unit 2 Baseline CDF QCNPS Unit 2 Baseline LERF Source Contribution Source Contribution Internal Events PRA 3.9E-06 Internal Events PRA 2.4E-07 Fire PRA 4.3E-05 Fire PRA 3.5E-06 Seismic 4.3E-06 Seismic 2.0E-06 No significant No significant Other External Events Other External Events contribution contribution Total CDF 5.1E-05 Total LERF 5.7E-06 E5-2
| |
| | |
| License Amendment Request Enclosure 5 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF)
| |
| As demonstrated in Table E5-1, the total CDF and total LERF are within the guidelines set forth in RG 1.174 and support small changes in risk that may occur during RICT entries following TSTF-505 implementation. Therefore, QCNPS TSTF-505 implementation is consistent with NEI 06-09 guidance. There will be a proceduralized check of the overall PRA results against the RG 1.174 thresholds in the PRA model update procedures.
| |
| The base internal events model for QCNPS is based on Unit 1 (Reference [5]), however the Unit 2 results are similar to the Unit 1 results. Therefore, the Unit 2 baseline CDF and LERF are taken from the Unit 1 specific model. The seismic LERF values listed in Table E5-1 are listed for the containment inerted state. For further explanation on seismic penalties in inerted vs de-inerted states, consult Enclosure 4.
| |
| : 3. References
| |
| [1] Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines',"
| |
| dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| [2] Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)
| |
| Guidelines," Revision 0, dated October 12, 2012 (ADAMS Accession No. ML12286A322)
| |
| [3] Regulatory Guide 1.174, "An Approach For Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, dated November 2002 (ADAMS Accession No. ML023240437)
| |
| [4] Regulatory Guide 1.174, "An Approach For Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, dated January 2018 (ADAMS Accession No. ML17317A256)
| |
| [5] QC-ASM-07, "2021 Fire PRA Completion Application-Specific (ASM) Notebook,"
| |
| Revision 0, dated October 2021
| |
| [6] QC-ASM-08, "Application-Specific Model (ASM) for Improving the Quantification Speed of the Fire PRA Model)," Revision 0, dated August 2022
| |
| [7] QC-MISC-037, "Seismic CDF and LERF Estimates for the Quad Cities TSTF-505 (RICT)
| |
| Program," Revision 0, dated December 2021 E5-3
| |
| | |
| ENCLOSURE 6 License Amendment Request Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Justification of Application of At-Power PRA Models to Shutdown Modes This enclosure is not applicable to the Quad Cities Nuclear Power Station submittal. Constellation is proposing to apply the Risk Informed Completion Time Program only in Modes 1 and 2 and not in the shutdown Modes.
| |
| | |
| ENCLOSURE 7 License Amendment Request Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" PRA Model Update Process
| |
| | |
| License Amendment Request Enclosure 7 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 PRA Model Update Process
| |
| : 1. Introduction Section 4.0, Item 8 of the U.S. Nuclear Regulatory Commission (NRC) Final Safety Evaluation (Reference [1]) for NEI 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines" (Reference [2]),
| |
| requires that the license amendment request (LAR) discuss the programs and procedures that assure the PRA models which support the RMTS are maintained consistent with the as-built/as-operated plant.
| |
| This enclosure describes the fleet-level administrative controls and procedural processes applicable to the configuration control of PRA models used to support the Risk Informed Completion Time (RICT) Program, which ensure that these models reflect the as-built/as-operated plant. Procedures and processes already implemented at other Constellation Energy Generation, LLC (CEG) sites will be used at Quad Cities Nuclear Power Station (QCNPS) where applicable. Prior to implementation, plant changes, including physical modifications and procedure revisions, will be identified and reviewed for the potential to impact the PRA models per ER-AA-600-1015, 'Full-Power Internal Events (FPIE) PRA Model Update,"
| |
| (Reference [3]) and ER-AA-600-1061, "Fire PRA Model Update and Control" (Reference [4]).
| |
| The configuration risk management program (CRMP) will ensure these plant changes are incorporated into the PRA models, as appropriate. This process includes discovered conditions associated with the PRA models, which will be addressed by the applicable site Corrective Action Program (CAP).
| |
| Should a plant change or a discovered condition be identified that has a significant impact to the RICT Program calculations as defined by the above procedures, an unscheduled update of the PRA model will be implemented. Otherwise, the PRA model change is incorporated into a subsequent periodic model update. Such pending changes are considered when evaluating other changes until they are fully implemented into the PRA models. Periodic updates are typically performed every two refueling cycles.
| |
| : 2. PRA Model Update Process Internal Event, Internal Flood, and Fire PRA Model Maintenance and Update The fleet risk management process ensures that the applicable PRA models used for the RICT Program reflect the as-built/as-operated plant for QCNPS. The PRA configuration control process delineates the responsibilities and guidelines for updating the full power internal events, internal flood, and fire PRA models, and includes both periodic and unscheduled PRA model updates.
| |
| The process includes provisions for monitoring potential impact areas affecting the technical elements of the PRA models (e.g., due to plant changes, plant/industry operational experience, or errors or limitations identified in the model), assessing the individual and cumulative risk impact of unincorporated changes, and controlling the model and necessary computer files, including those associated with the Real-Time Risk (RTR) model.
| |
| E7-2
| |
| | |
| License Amendment Request Enclosure 7 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 PRA Model Update Process Changes that are considered an upgrade per the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA standard receive a peer review focused on those aspects of the PRA model that represent the upgrade.
| |
| Review of Plant Changes for Incorporation into the PRA Model
| |
| : 1) Plant changes or discovered conditions are reviewed for potential impact to the PRA models, including the RTR model and the subsequent risk calculations which support the RICT Program (NEI 06-09, Section 2.3.4, Items 7.2 and 7.3, and Section 2.3.5, Items 9.2 and 9.3).
| |
| : 2) Plant changes that meet the criteria defined in References 3 and 4 (including consideration of the cumulative impact of other pending changes) will be incorporated in the applicable PRA model(s). Otherwise, the change is assigned a priority and incorporated during a subsequent periodic update consistent with procedural requirements. (NEI 06-09, Section 2.3.5, Item 9.2)
| |
| : 3) PRA updates for plant changes are performed at least once every two refueling cycles (NEI 06-09, Section 2.3.4, Item 7.1, and Section 2.3.5, Item 9.1).
| |
| : 4) If a PRA model change is required for the RTR model but cannot be immediately implemented for a significant plant change or discovered condition, either of the following actions can be used to satisfy NEI 06-09, Section 2.3.5, Item 9.3.
| |
| a) Interim analyses to address the expected risk impact of the change will be performed. In such a case, these interim analyses become part of the RICT Program calculation process until the plant changes are incorporated into the PRA model during the next update.
| |
| OR b) Appropriate administrative restrictions on the use of the RICT Program for extended Completion Times are put in place until the model changes are completed.
| |
| : 3. References
| |
| [1] Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines',"
| |
| dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| [2] Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)
| |
| Guidelines," Revision 0, dated October 12, 2012 (ADAMS Accession No. ML12286A322)
| |
| E7-3
| |
| | |
| License Amendment Request Enclosure 7 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 PRA Model Update Process
| |
| [3] ER-AA-600-1015, "FPIE PRA Model Update," Revision 20, dated February 2020
| |
| [4] ER-AA-600-1061, "Fire PRA Model Update and Control," Revision 7, dated February 2020 E7-4
| |
| | |
| ENCLOSURE 8 License Amendment Request Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Attributes of the Real-Time Risk Model
| |
| | |
| License Amendment Request Enclosure 8 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Attributes of the Real-Time Risk Model
| |
| : 1. Introduction Section 4.0, Item 9 of the U.S. Nuclear Regulatory Commission (NRC) Final Safety Evaluation (Reference [1]) for NEI 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines" (Reference [2]),
| |
| requires that the license amendment request (LAR) provide a description of Probabilistic Risk Assessment (PRA) models and tools used to support the RMTS. This includes identification of how the baseline PRA model is modified for use in the configuration risk management program (CRMP) tools, quality requirements applied to the PRA models and CRMP tools, consistency of calculated results from the PRA model and the CRMP tools, and training and qualification programs applicable to personnel responsible for development and use of the CRMP tools.
| |
| NEI 06-09, Revision 0-A, uses the term CRMP for the program controlling the use of RMTS.
| |
| This term is also used to designate the program implementing 10 CFR 50.65(a)(4) and the monitoring program for other risk-informed LARs. This enclosure will use the term Risk Informed Completion Time (RICT) Program to indicate the program required by NEI 06-09, Revision 0-A, in lieu of CRMP. Item 9 should also confirm that the RICT Program tools can be readily applied for each Technical Specification (TS) Limiting Condition for Operation (LCO) within the scope of the plant-specific submittal.
| |
| This enclosure describes the necessary changes to the peer-reviewed baseline PRA models for use in the Real-Time Risk (RTR) tool to support the RICT Program. The process employed to adapt the baseline models is demonstrated to:
| |
| a) Preserve the core damage frequency (CDF) and large early release frequency (LERF) quantitative results, b) Maintain the quality of the peer-reviewed PRA models, and c) Correctly accommodate changes in risk due to configuration-specific considerations.
| |
| Quality controls and training programs applicable for the RICT Program are also discussed in this enclosure.
| |
| : 2. Translation of Baseline PRA Model for Use in Configuration Risk The baseline PRA models for internal events, including internal flood and internal fire, are the peer-reviewed models. These models are updated when necessary to incorporate plant changes to reflect the as-built/as-operated plant. The internal flood model is integrated into the internal events model. These models will be used in the RICT Program. The models may be optimized for quantification speed but are verified to provide the same result as the baseline models in accordance with approved procedures.
| |
| The RTR tool will be used to facilitate all configuration-specific risk calculations and support the RICT Program implementation. The PRA models utilize system initiator event fault trees which explicitly capture equipment unavailabilities. Therefore, no adjustment to initiating event frequencies are required within the RTR tool.
| |
| E8-2
| |
| | |
| License Amendment Request Enclosure 8 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Attributes of the Real-Time Risk Model The baseline PRA models are modified as follows for use in configuration risk calculations:
| |
| Unit availability factor is set to 1.0 (unit available)
| |
| Maintenance unavailability is set to zero/false unless unavailable due to the configuration Mutually exclusive combinations, including normally disallowed maintenance combinations, are adjusted to allow accurate analysis of the configuration For systems where some trains are in service and some in standby, the RTR tool addresses the actual plant configuration including defining in service trains as needed Impact of outside temperatures on system requirements like seasonal service water pumps will be evaluated and adjusted with flags in the CRMP model as needed No changes in success criteria based on the time in the core operating cycle are required Systems with shared components across units which are credited in the RTR models are represented in the RTR tools for both units simultaneously, reflecting availability or unavailability of the shared system to each unit based on the actual plant configuration. For a RICT program entry, the unit RTR tool will reflect the actual configuration of the plant, including availability or unavailability of shared systems and components.
| |
| The configuration risk software is designed to quantify the specific configuration for internal events, including internal flooding and fire, and includes the seismic risk contribution when calculating the Risk Management Action Time (RMAT) and RICT. Full quantifications will be used for each configuration. Pre-solved cutsets will be limited to results for specific configurations. For configurations without pre-solved cutsets the model will be quantified to produce cutsets for the previously unanalyzed configuration. If there are any changes in the underlying PRA, the PRA results database in PARAGON will be updated in accordance with the RTR update procedure (Reference [7]). The unique aspect of the configuration risk software for the RICT program is the quantification of fire risk and the inclusion of the seismic risk contribution. The other adjustments above are those used for the evaluation of risk under the 10CFR 50.65(a)(4) program.
| |
| The Quad Cities Nuclear Power Station (QCNPS) PRA calculates common cause basic event (CCBE) probabilities from alpha factors and places the basic events under appropriate gates in the fault tree.
| |
| Adjustments to the common cause failure (CCF) grouping or CCF probabilities are not necessary when a component is taken out-of-service for preventative maintenance (PM) because:
| |
| The in-service components are not subject to increases in common cause probabilities since the component is not out-of-service for a reason subject to a potential CCF.
| |
| CCF relationships are retained for the remaining in-service components.
| |
| The net failure probability for the in-service components includes the CCF contribution of the out-of-service component.
| |
| E8-3
| |
| | |
| License Amendment Request Enclosure 8 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Attributes of the Real-Time Risk Model As described in Regulatory Guide (RG) 1.177 (Reference [6]), Section A-1.3.2.2, the CCF term should be treated differently when a component is taken down for PM than for failure. For PMs, the common cause factor is changed so that the model represents the unavailability of the remaining component. In the example provided in RG 1.177 for a two-train system, the CCF event can be set to zero for PMs. This is done so that the model represents the unavailability of the remaining component, and not the common cause multiplier. The QCNPS approach is conservative in that for a two-train system, the CCF event is retained for the component removed from service. Likewise, for systems with three or more trains, the CCF events that are related to the out-of-service component are retained.
| |
| The Vogtle RICT Safety Evaluation (SE) (Reference [5]) describes the Vogtle approach for modeling common cause events with planned inoperability: "For planned inoperability, the licensee sets the appropriate independent failure to 'true' and makes no other changes while calculating a RICT." The QCNPS approach is the same as this Vogtle approach.
| |
| It is recognized that other modifications could be made to CCF factors for planned maintenance, particularly for common cause groups of three or more components. For example, the Vogtle RICT Amendment SE (Reference [5]) identifies a possible planned maintenance CCF modification to "modify all the remaining basic event probabilities to reflect the reduced number of redundant components."
| |
| Like Vogtle, the QCNPS CCF approach is a straightforward simplification that has inherent uncertainties. In the context of modifying CCF basic events for PMs, the Vogtle SE (Reference [5]) states the following:
| |
| "The NRC staff also notes that common cause failure probability estimates are very uncertain and retaining precision in calculations using these probabilities will not necessarily improve the accuracy of the results. Therefore, the NRC staff concludes that the licensee's method is acceptable because it does not systematically and purposefully produce non-conservative results and because the calculations reasonably include common cause failures consistent with the accuracy of the estimates."
| |
| The QCNPS approach for CCF during PMs is the same as the Vogtle approach; therefore, adjusting the common cause grouping is not necessary for PMs. However, if a numeric adjustment is performed, the RICT calculation shall be adjusted to numerically account for the increased possibility of CCF as specified in Section A-1.3.2.1 of Appendix A of RG 1.177.
| |
| For emergent conditions where the extent of condition (EOC) is not completed prior to entering the RMATs or the EOC cannot rule out the potential for CCF, common cause Risk Management Actions (RMAs) are expected to be implemented to mitigate CCF potential and impact, in accordance with Constellation Energy Generation, LLC (CEG) procedures. This is in line with the guidance of NEI 06-09 and precludes the need to adjust CCF probabilities. However, if a numeric adjustment is performed, the RICT calculation shall be adjusted to numerically account for the increased possibility of CCF as specified in Section A-1.3.2.1 of Appendix A of RG 1.177.
| |
| E8-4
| |
| | |
| License Amendment Request Enclosure 8 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Attributes of the Real-Time Risk Model
| |
| : 3. Quality Requirements and Consistency of PRA Model and Configuration Risk Tools The approach for establishing and maintaining the quality of the PRA models, including the configuration risk model, includes both a PRA maintenance and update process (described in ), and the use of self-assessments and independent peer reviews (described in ).
| |
| The information provided in Enclosure 2 demonstrates that the sites internal event, internal flood, and internal fire PRA models reasonably conform to the associated industry standards endorsed by RG 1.200 (Reference [3]). This information provides a robust basis for concluding that the PRA models are of sufficient quality for use in risk-informed licensing actions.
| |
| For maintenance of an existing configuration risk model, changes made to the baseline PRA model in translation to the configuration risk model will be controlled and documented. Every PRA model of record (MOR) update results in an update to the RTR model in accordance with the internal events and fire PRA update procedures (Reference [7]). An acceptance test is performed after every configuration risk model update. This testing also verifies correct mapping of plant components to the basic events in the configuration risk model. The RTR model documentation includes changes made to the MOR model files to work with the RTR model software (e.g., quantification settings) along with verification that the zero maintenance results are consistent between the RTR and PRA. In addition, the RTR update for the MOR includes quantifying the RTR model for representative maintenance configurations and examining the results for appropriateness. These actions are procedurally controlled.
| |
| : 4. Training and Qualification The PRA staff is responsible for development and maintenance of the configuration risk model.
| |
| Operations and Work Control staff will use the configuration risk tool under the RICT Program.
| |
| PRA staff and Operations are trained in accordance with a program using National Academy for Nuclear Training (ACAD) documents, which is accredited by the Institute of Nuclear Power Operations (INPO).
| |
| : 5. Application of the CRMP Tool to the RICT Program Scope The PARAGON software will be used to facilitate all configuration-specific risk calculations and support RICT Program implementation. QCNPS has its own PARAGON model specifically designed to support implementation of RMTS. PARAGON will permit the user to evaluate all plant configurations using appropriate mapping of equipment to PRA basic events. The equipment in the scope of the RICT Program will be able to be evaluated in the appropriate PRA models. The RICT Program will meet RG 1.174 (Reference [4]) and CEG software quality assurance requirements.
| |
| E8-5
| |
| | |
| License Amendment Request Enclosure 8 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Attributes of the Real-Time Risk Model
| |
| : 6. References
| |
| [1] Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines',"
| |
| dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| [2] Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)
| |
| Guidelines," Revision 0, dated October 12, 2012 (ADAMS Accession No. ML12286A322)
| |
| [3] Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, dated March 2009 (ADAMS Accession No. ML090410014)
| |
| [4] Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, dated January 2018 (ADAMS Accession No. ML17317A256)
| |
| [5] Letter from G.E. Miller (U.S. NRC) to J.J. Hutto (Southern Company), "Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments Regarding Implementation of Topical Report Nuclear Energy Institute NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specification (RMTS) Guidelines, Revision 0-A (CAC NOS. ME9555 and ME9556)," dated August 8, 2017 (ADAMS Accession No. ML15127A669)
| |
| [6] Regulatory Guide 1.177, "Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," Revision 2, dated January 2021 (ADAMS Accession No. ML20164A034)
| |
| [7] ER-AA-600-1016, Configuration Risk Model Update, Revision 14, dated July 2022 E8-6
| |
| | |
| ENCLOSURE 9 License Amendment Request Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Key Assumptions and Sources of Uncertainty
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty
| |
| : 1. Introduction The purpose of this enclosure is to disposition the impact of Probabilistic Risk Assessment (PRA) modeling epistemic uncertainty for the Risk Informed Completion Time (RICT) Program as part of RICT Technical Specification Task Force (RITSTF) Initiative 4b. Nuclear Energy Institute (NEI) Topical Report NEI 06-09-A (Reference [1]) Section 2.3.4, item 10 requires an evaluation to determine insights that will be used to develop Risk Management Actions (RMAs) to address these uncertainties. The baseline internal events PRA and fire PRA (FPRA) models document assumptions and sources of uncertainty and these were reviewed during the model peer reviews. The approach taken is, therefore, to review these documents to identify the items which may be directly relevant to the RICT Program calculations, to perform sensitivity analyses where appropriate, to discuss the results and to provide dispositions for the RICT Program.
| |
| The epistemic uncertainty analysis approach described below applies to the internal events PRA and any epistemic uncertainty impacts that are unique to FPRA are also addressed. In addition, Topical Report NEI 06-09-A requires that the uncertainty be addressed in RICT Program Configuration Risk Management Program (CRMP), otherwise referred to as the Real-Time Risk (RTR), tools by consideration of the translation from the PRA model to the RTR model. The RTR model, also referred to as the PARAGON model, discussed in Enclosure 8 includes internal events, flooding events and fire events. The model translation uncertainties evaluation and impact assessment are limited to new uncertainties that could be introduced by application of the RTR tool during RICT Program calculations.
| |
| In addition, generic level 2 sources of uncertainty and assumptions from EPRI 1026511 (Reference [2]) were also evaluated. No additional key sources of model uncertainty were identified from that review of Level 2 generic uncertainties.
| |
| : 2. Diverse and Flexible Mitigating Strategies (FLEX) in the PRA Models The purpose of the QCNPS FLEX equipment is to provide independent means capable of mitigating a simultaneous loss of all AC power and loss of normal access to the ultimate heat sink resulting from a beyond design basis external event (BDBEE) by providing adequate capability to maintain or restore core cooling, containment cooling, and Spent Fuel Pool (SFP) cooling capabilities to all units on site.
| |
| Through procedural direction, the use of FLEX equipment is limited to extended loss of AC power (ELAP) conditions (e.g., station blackout (SBO)).
| |
| At QCNPS, FLEX equipment consists of the following:
| |
| Five 500KW portable diesel generators, three located in the FLEX Building and two backup generators located in the N+1 Building.
| |
| o Of the generators in the FLEX Building, one is dedicated for Unit 1 supply power to Buses 18/19, one is dedicated for Unit 2 supply power for Buses 28/29, and the third is used for supply power to the Deep Well Pump.
| |
| The Seismic Deep Well Pump is located east of the 13.8kV switchyard and provides a diverse power independent source of water to makeup to E9-2
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty the reactor vessel and fuel pools as well as provide for containment cooling and radioactive release mitigation.
| |
| o The FLEX Building also houses a FLEX Yanmar portable diesel generator used for smaller temporary loads (i.e., Control Room ventilation, Battery Room ventilation) and an ADS portable DC diesel generator used to provide DC power to operate at least two ADS valves.
| |
| Three Godwin pumps are used to supply water to the Reactor, Torus, and SFP in the event the Deep Well Pump is no longer functional or during a severe accident condition.
| |
| Two Godwin pumps are stored in the FLEX Building while the third is stored in the N+1 Building.
| |
| Associated temporary and installed piping, valves, and instrumentation.
| |
| QCNPS also has at its disposal two other mechanisms (systems) for addressing conditions which may arise due to ELAP, namely, the Hardened Containment Vent System (HCVS) and BlackStarTech (also referred to as BlackStart). Because operation of either of these two systems is subject to the same entry criteria as FLEX (e.g., ELAP (SBO)), HCVS and BlackStart are often included in discussions and evaluations which include FLEX.
| |
| The HCVS provides a vent pipe, motive power for vent valves, a purge system, and a radiation and temperature monitoring system to enable venting of the suppression chambers during and following an ELAP (FLEX) or Severe Accident Water Addition (SAWA) event. This is a unit specific vent path directly from the containment vent header to an elevated release point and is powered by the 125 VDC batteries for valve control and instrumentation for up to 24 hours.
| |
| BlackStart is a collection of rapid deployment power supplies and attendant equipment for use during an extended station blackout. BlackStart is comprised of portable battery, generator, inverter, and various voltage AC and DC power supplies, and can supply a wide range of loads from 24 VDC to 250 VDC and 110 VAC to 240 VAC.
| |
| Batteries provide rapid power restoration and portable liquified petroleum gas (LPG, i.e.,
| |
| propane) generators provide long-term power. The goal of BlackStart is rapid re-establishment of a reactor water injection source, reactor pressure control, and vital instrumentation needed to monitor the health of the reactor core and containment structures. BlackStart procedures are written to prioritize restoring injection and pressure control, followed by restoring instrumentation and control functions using the BlackStart batteries. Once the vital systems are restored, the focus shifts to providing LPG generator connections to extend the coping time.
| |
| The QCNPS PRA models include FLEX, HCVS, and BlackStart. However, due to the procedurally directed entry conditions (e.g., ELAP), these three systems are credited as available accident mitigation features within the PRA only if an SBO occurs; SBOs are rare events, and therefore the impacts of FLEX, HCVS, and BlackStart upon core damage frequency and large early release frequency are limited.
| |
| In the base PRA model, FLEX equipment failure rates are estimated by doubling the failure rates for similar component types (for example, failure rates for standard emergency diesel generators are multiplied by a factor of two as an estimate for the failure rates of the FLEX E9-3
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty 500 KW diesel generators). FLEX-related human error probabilities (HEPs) use screening values. HCVS failure rates are based on a combination of the factor of two approach for some components (e.g., HCVS 24 VDC battery charger) and rates for like-components from NUREG/CR-6928 (updated with plant-specific experience when available), such as for air-operated butterfly valves. HCVS HEPs also employ screening values. Finally, BlackStart failure rates are based on NUREG/CR-6928 as well, while BlackStart HEPs are calculated using detailed human reliability analysis (HRA) techniques.
| |
| For the uncertainty analyses discussed in this enclosure, credit for FLEX, HCVS, and BlackStart in the PRA was removed, both individually (one at a time) and collectively (all three simultaneously). Because PRA credit is available only during SBO conditions, and because the frequency of SBO events at QCNPS is assessed to be very low, there is an insignificant impact on CDF and LERF when FLEX, HCVS, and BlackStart are unavailable in response to an FPIE or fire initiating event. The total CDF increase for FPIE plus fire initiating events given simultaneous unavailability of FLEX, HCVS, and BlackStart is essentially imperceptible, while for LERF the increase is less than 1% (Reference [3]). When assessing the impact of no credit on select RICT cases, the reduction in RICT days is either zero (0), or, in non-zero RICT cases, approximately 0.1 days (Reference [3]).
| |
| : 3. Assessment of Internal Events PRA Epistemic Uncertainty Impacts To identify key sources of uncertainty, the Internal Events baseline PRA model uncertainty report was developed, based on the guidance in NUREG-1855 (Reference [3]) and EPRI 1016737 (Reference [4]). As described in NUREG-1855, sources of uncertainty include "parametric" uncertainties, "modeling" uncertainties, and "completeness" (or scope and level of detail) uncertainties.
| |
| Parametric uncertainty was addressed as part of the Quad Cities Nuclear Power Station (QCNPS) baseline PRA model quantification. The parametric uncertainty evaluation for the Internal Events PRA model is documented in Appendix B of the Summary Notebook (Reference [5]).
| |
| Modeling uncertainties are considered in both the base PRA and in specific risk-informed applications. Assumptions are made during the PRA development to address a particular modeling uncertainty because there is not a single definitive approach. Plant-specific assumptions made for each of the QCNPS Internal Events PRA technical elements are noted in the Summary Notebook (Reference [5]). The Internal Events PRA model uncertainties evaluation considers the modeling uncertainties for the base PRA by identifying assumptions, determining if those assumptions are related to a source of modeling uncertainty and characterizing that uncertainty, as necessary. The Electric Power Research Institute (EPRI) compiled a listing of generic sources of modeling uncertainty to be considered for each PRA technical element (Reference [4]), and the evaluation performed for QCNPS considered each of the generic sources of modeling uncertainty as well as the plant-specific sources.
| |
| Completeness uncertainty addresses scope and level of detail. Uncertainties associated with scope and level of detail are documented in the PRA but are only considered for their impact on a specific application. No specific issues of PRA completeness have been identified relative to the RICT application, based on the results of the Internal Events PRA and Fire PRA peer reviews.
| |
| E9-4
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty The detailed evaluation of assumptions and sources of uncertainty for the FPIE Model is included in Reference [6]. The summary evaluation of those FPIE assumptions and sources of uncertainty determined to be "key" for this application are shown in Table E9-1 below.
| |
| : 4. Assessment of Supplementary FPRA Epistemic Uncertainty Impacts The purpose of the following discussion is to address the epistemic uncertainty in the QCNPS Fire PRA (FPRA). The FPRA model includes various sources of uncertainty that exist because there is both inherent randomness in elements that comprise the FPRA, and because the state of knowledge in these elements continues to evolve. The development of the FPRA was guided by NUREG/CR-6850 (Reference [7]). The FPRA model used consensus models described in NUREG/CR-6850. Enclosure 2 provides a detailed discussion of the Peer Review F&Os and resolutions.
| |
| In order to identify key sources of uncertainty for the RICT Program Application, an evaluation of Fire PRA model uncertainty was performed, based on the guidance in NUREG-1855 (Reference [3]) and Electric Power Research Institute (EPRI) report 1026511 (Reference [2]).
| |
| As stated in Section 1.3 of NUREG-1855:
| |
| Although the guidance in the this [sic] report does not currently address all sources of uncertainty, the guidance provided on the uncertainty identification and characterization process and on the process of factoring the results into the decision making is generic and independent of the specific source of uncertainty. Consequently, the guidance is applicable for sources of uncertainty in PRAs that address at-power and low power and shutdown operating conditions, and both internal and external hazards.
| |
| NUREG-1855 also describes an approach for addressing sources of model uncertainty and related assumptions. It states:
| |
| A source of model uncertainty exists when (1) a credible assumption (decision or judgment) is made regarding the choice of the data, approach, or model used to address an issue because there is no consensus and (2) the choice of alternative data, approaches or models is known to have an impact on the PRA model and results. An impact on the PRA model could include the introduction of a new basic event, changes to basic event probabilities, change in success criteria, or introduction of a new initiating event. A credible assumption is one submitted by relevant experts and which has a sound technical basis. Relevant experts include those individuals with explicit knowledge and experience for the given issue. An example of an assumption related to a source of model uncertainty is battery depletion time. In calculating the depletion time, the analyst may not have any data on the time required to shed loads and thus may assume (based on analyses) that the operator is able to shed certain electrical loads in a specified time.
| |
| Section 2.1.3 of NUREG-1855 defines consensus model as:
| |
| A consensus model is a model that has a publicly available published basis and has been peer reviewed and widely adopted by an appropriate stakeholder group. In E9-5
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty addition, widely accepted PRA practices may be regarded as consensus models.
| |
| Examples of the latter include the use of the constant probability of failure on demand model for standby components and the Poisson model for initiating events. For risk-informed regulatory decisions, the consensus model approach is one that the NRC has used or accepted for the specific risk-informed application for which it is proposed.
| |
| Modeling uncertainties are considered in both the base Fire PRA and in specific risk-informed applications. Assumptions are made during the FPRA development as a way to address a particular modeling uncertainty because there is not a single definitive approach. Plant-specific assumptions made for each of the QCNPS FPRA technical elements are noted in the Uncertainty and Sensitivity Analysis Notebook (Reference [8]). Reference [6] contains discussion of the sources of uncertainty identified for each task of NUREG/CR-6850 (Reference [7]).
| |
| Completeness uncertainty addresses scope and level of detail. Uncertainties associated with scope and level of detail are documented in the Fire PRA but are only considered for their impact on a specific application. No specific issues of PRA completeness have been identified relative to the RICT application, based on the results of the Internal Events PRA and Fire PRA peer reviews.
| |
| EPRI compiled a listing of generic sources of modeling uncertainty to be considered for each Fire PRA technical element in EPRI report 1026511 (Reference [2]). Based on following the methodology in EPRI 1026511 for a review of sources of uncertainty, the impact of potential sources of uncertainty on the PRA or applications is discussed in Reference [6], which identifies those potential sources that may be key sources of uncertainty for the RICT program. Note that for RICT applications, RMAs will be developed when appropriate using insights from the Fire PRA model results specific to the configuration. This will include a review for impacts due to the key sources of uncertainty.
| |
| A detailed evaluation of assumptions and sources of uncertainty for the FPRA Model is included in Reference [6]. The summary evaluation of those FPRA assumptions and sources of uncertainty determined to be "key" for this application are shown in Table E9-1 below.
| |
| E9-6
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty
| |
| : 4. Assessment of FPIE and FPRA Model Key Assumptions and Uncertainty Based on the discussions in Sections 2 and 3 above, key assumptions and sources of uncertainty determined to be key for this application are identified and dispositioned in Table E9-1.
| |
| Table E9-1: Identified Potential Key Assumptions and Uncertainties - Internal Hazards (Internal Events, Flood, and Fire)
| |
| Sources of Assumption/ Model Sensitivity and RICT Impact Uncertainty Disposition (RICT)
| |
| Internal Events and Internal Flood (IE/IF) Model Core Cooling Success FW / Condensate / SBCS and A sensitivity analysis was Following Containment CRD are credited for success performed that increased Failure or Venting Through after containment failure, but an the conditional probability Non-Hard Pipe Vent Paths additional basic event by a factor of 10 (to (1CNPVDWRUPT--R-- large DW 6.0E-1) that a large drywell Following containment containment failure causes loss of failure would result in loss failure, injection from CRD injection) is included that of the feedwater, and FW/Condensate could represents the likelihood that the condensate, SBCS, and still be maintained, but if a containment failure size and CRD injection capabilities.
| |
| large containment failure location disrupts the capability of occurs, injection paths may FW/Condensate/SBCS and CRD The results demonstrate be disrupted leading to loss to inject. that FPIE PRA and FPRA of these external sources. CDF and LERF results are This failure probability is sensitive to the failure based on a detailed probability associated with structural analysis of the failure of all of the selected Mark I containment design injection systems following and large scale ultimate a large drywell failure.
| |
| failure testing of steel However, a factor of 10 containments. increase in the conditional failure probability, applied to all of the selected injection systems, is not considered credible.
| |
| HPCI Room Cooling HVAC dependencies for HPCI are The requirement for room included given gland seal cooling for various HPCI HPCI room cooling is condenser failure. mission times, with and supplied by DGCW1. It without failure of the gland requires manual startup seal condenser, is action. No room cooling is identified as a candidate required for HPCI mission source of model time as long as there is no uncertainty.
| |
| E9-7
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty Table E9-1: Identified Potential Key Assumptions and Uncertainties - Internal Hazards (Internal Events, Flood, and Fire)
| |
| Sources of Assumption/ Model Sensitivity and RICT Impact Uncertainty Disposition (RICT) gland seal condenser failure. For gland seal A sensitivity analysis was failures, the HPCI system is performed that increased assigned failure directly. the failure probability for the HPCI gland seal hotwell pump failing to start by about a factor of 100 (to 1.0E-1).
| |
| The results demonstrate that FPIE PRA CDF and LERF have little sensitivity to the failure probability associated with failure of HPCI room cooling.
| |
| However, a factor of about 100 increase in the pumps failure probability is conservative and not consistent with observed behavior.
| |
| Digital Feedwater Control Reliability values from vendor For this sensitivity Failure Probabilities studies demonstrating that the analysis, the failure system performance would result probability for the digital There are model in less than 0.1 transients per feedwater controller failing uncertainties associated year are used for the key to control feedwater such with modeling digital components of the system. that the RPV overfills and systems, such as those floods the steam line was related to determining the Basic events representing the increased by a factor of failure modes of these reliability values for the auto level 100.
| |
| systems and components. controller, the field buses, false signal from the redundant The results demonstrate reactivity control system, and that FPIE PRA CDF and false signal from the Level 8 trip LERF are not sensitive to system are included in the system the failure probability logic model. associated with failure of the digital feedwater controller to stop feedwater prior to vessel overfill.
| |
| E9-8
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty Table E9-1: Identified Potential Key Assumptions and Uncertainties - Internal Hazards (Internal Events, Flood, and Fire)
| |
| Sources of Assumption/ Model Sensitivity and RICT Impact Uncertainty Disposition (RICT)
| |
| Instrument Air (IA) System It is assumed the containment The recovery probability Recovery (Containment vent valves cannot be opened by assigned for recovering Vent Valve Dependency on local manipulation of the valves or instrument air is Air) their air operators. They require considered reasonable instrument air to provide the force and is supported by data.
| |
| The containment vent to open the containment vent However, use of this value valves do not have AOVs. This requires instrument could lead to a slightly accumulator backups to air availability to fulfill the optimistic assessment of provide a method of containment vent function. For containment vent success.
| |
| successful venting given a some low probability sequences, loss of IA scenario. instrument air is not available due For this sensitivity Currently, the model credits to system failures. However, the analysis, the probability of IA recovery at 24 hours to model uses data from NSAC-161 failing to recover balance restore the hard pipe vent (Reference [9]) to provide a basis of plant systems (including path. to support the restoration of instrument air) after IA instrument air or its support failure due to random systems. causes was increased to 1.0.
| |
| The results demonstrate that FPIE PRA CDF and LERF are relatively insensitive to the failure probability associated with failure to recover instrument air (balance of plant, as well as specific to support of venting).
| |
| FLEX and Hardened The FLEX component failure The HEPs and equipment Containment Vent System rates can be represented by using reliabilities used for FLEX (HCVS) human error the failure rates of "like- may be underestimated probabilities (HEPs) and components" (e.g., emergency given the current state of Equipment Failure Rates diesel generator (EDG)) as knowledge about FLEX.
| |
| surrogates (e.g., for the FLEX For the QCNPS IE PRA diesel generators). The HEPs For this sensitivity case, model there were no can be represented using credit for FLEX was industry-approved data screening values (ranging from completely removed from sources for FLEX 1E-3 to 0.5). FLEX component the model through use of a equipment reliability. The failures are estimated based on FLAG event.
| |
| QCNPS PRA models FLEX "like-components" by increasing component failures and that like-components failure rate E9-9
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty Table E9-1: Identified Potential Key Assumptions and Uncertainties - Internal Hazards (Internal Events, Flood, and Fire)
| |
| Sources of Assumption/ Model Sensitivity and RICT Impact Uncertainty Disposition (RICT) human failure events by a factor of two. Human error The results demonstrate associated with failure to probabilities employ screening that FPIE PRA CDF and align the FLEX equipment. values. LERF essentially have no sensitivity to the availability of FLEX. This result is driven by the fact that the use of FLEX at the QCNPS is constrained to situations declared as Extended Loss of AC Power (ELAP) scenarios (including station blackout (SBO)). These scenarios are rare, and thus there are few sequences in the PRA for which FLEX is credited. Removing that credit therefore has very little impact on the quantified results.
| |
| BlackStarTech (BST) Use of BST has the same The base model includes portable carts equipment constraints as FLEX and HCVS credit for the BST portable and human action reliability (e.g., ELAP). The BST equipment (which includes component failure rates can be the capability to supply For the QCNPS PRA represented by using the failure specific AC and DC power model there were no rates of "like-components" (e.g., to select components).
| |
| industry-approved data batteries and battery chargers) as Equipment failure sources for BST equipment surrogates for BST equipment. probabilities are based on reliability. The QCNPS The HEPs can be represented similar components (e.g.,
| |
| PRA models BST using standard HRA techniques batteries), and human component failures and (as employed for other human error probabilities are human failure events failure events throughout the calculated using detailed associated with failure to PRA). human reliability analysis align the BST equipment. techniques.
| |
| For this sensitivity case, credit for BST was completely removed from the model through use of a FLAG event. The results demonstrate that FPIE E9-10
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty Table E9-1: Identified Potential Key Assumptions and Uncertainties - Internal Hazards (Internal Events, Flood, and Fire)
| |
| Sources of Assumption/ Model Sensitivity and RICT Impact Uncertainty Disposition (RICT)
| |
| PRA CDF and LERF have no sensitivity to the availability of BST. Use of BST has the same constraints as FLEX and HCVS (e.g., ELAP).
| |
| Fire Model Fire PRA component The Fire PRA assumes that, at a For this sensitivity case, selection involves the minimum, a plant trip occurs. equipment or cables selection of components to This is consistent with accepted assumed failed or credited be treated in the analysis in industry practice. The Fire PRA by assumed routing in the the context of fire initiating does not credit some equipment base FPRA were assumed events and mitigation. The or systems that are credited in the to always be available.
| |
| potential sources of full power internal events PRA.
| |
| uncertainty include those Systems are included based on The results demonstrate inherent in the internal an iterative process to include that FPRA results are events PRA model as that equipment that may be significant sensitive to changes when model provides the to the fire risk. the equipment is assumed foundation for the FPRA. available. However, a review of the sensitivity analysis results identified the reasons for the decrease in FPRA results were non-conservative given the contributing equipment is not expected to be available.
| |
| The scope of credited equipment and cables and assumed cable routing is based on reviews of the applicable systems and the PRA model.
| |
| Therefore, the scope credited equipment in the FPRA provides best estimate results.
| |
| E9-11
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty Table E9-1: Identified Potential Key Assumptions and Uncertainties - Internal Hazards (Internal Events, Flood, and Fire)
| |
| Sources of Assumption/ Model Sensitivity and RICT Impact Uncertainty Disposition (RICT)
| |
| The Human Error The Fire PRA includes The use of 1E-6 is Probabilities (HEPs) used conservative adjustments to the consistent with industry in the FPRA were adjusted HFEs to account for adverse guidance for FPRA and is to consider the additional impacts of fire events. The Fire an accepted practice.
| |
| challenges that may be PRA does not include credit for all present given a fire. The operator actions, including fire A sensitivity case was HEPs included the response actions. The Fire PRA performed for the base consideration of does not include credit for all FPRA using a minimum degradation or loss of instrument cues that may be joint HEP of 1E-5. The necessary cues due to fire. available. A minimum joint HEP results demonstrate that Given the methodology was applied for the HRA using a higher FPRA used, the impact of any dependency analysis. Applying a minimum joint HEP has a remaining uncertainties is minimum joint HEP may skew the slight impact on FPRA expected to be small. results by artificially increasing the results.
| |
| risk due to human failure events.
| |
| The HEPs are propagated in the parametric uncertainty evaluation based on the uncertainty parameters from the HRAC.
| |
| : 5. Translation (CRMP Model) Uncertainty Impact Assessment Incorporation of the baseline PRA models into the RTR model used for RICT Program calculations may introduce new sources of model uncertainty. Table E9-2 provides a description of the relevant model changes and dispositions of whether any of the changes made represent possible new sources of model uncertainty that must be addressed. Refer to for additional discussion on the RTR model.
| |
| Table E9-2: Assessment of Translation Uncertainty Impacts CRMP Model Part of Change and Model Impact on Model Disposition Assumptions Affected PRA model logic Fault tree The model, if restructured, Since the restructured structure may be logic model will be logically equivalent model will produce optimized to increase structure, and produce results comparable numerical solution speed. affecting comparable to the baseline results, this is not a both Internal PRA logic model. source of uncertainty for Events and the RICT Program.
| |
| Fire PRAs.
| |
| E9-12
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty Table E9-2: Assessment of Translation Uncertainty Impacts CRMP Model Part of Change and Model Impact on Model Disposition Assumptions Affected Incorporation of Calculation The addition of bounding Since this is a bounding seismic risk bias to of RICT and impacts for seismic events approach for addressing support RICT Risk has no impact on baseline seismic risk in the RICT Program risk Management PRA or RTR model. Program, it is not a calculations. Action Impact is reflected in source of translation Threshold calculation of all RICTs uncertainty, and RICT A conservative value (RMAT) and RMATs. Program calculations for the seismic delta within RTR. are not impacted, so no CDF is applicable. mandatory Risk Management Actions (RMAs) are required.
| |
| Set plant availability Typecode Since the RTR model This change is (Reactor Critical @AVAIL evaluates specific consistent with RTR tool Years Factor) basic configurations during at- practice; therefore, this event to 1.0. power conditions, the use change does not of a plant availability factor represent a source of less than 1.0 is not uncertainty, and RICT appropriate. This change Program calculations allows the RTR model to are not impacted, so no produce appropriate mandatory RMAs are results for specific at- required.
| |
| power configurations.
| |
| : 6. REFERENCES
| |
| [1] Nuclear Energy Institute (NEI) Topical Report (TR) 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"
| |
| Revision 0, dated October 12, 2012 (ADAMS Accession No. ML12286A322)
| |
| [2] EPRI TR-1026511, "Practical Guidance on the Use of PRA in Risk-Informed Applications with a Focus on the Treatment of Uncertainty," dated December 4, 2012 (publically available at https://www.epri.com/research/products/000000000001026511)
| |
| [3] NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," Revision 1, dated March 2017 (ADAMS Accession No. ML17062A466)
| |
| [4] EPRI TR-1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," dated December 19, 2008 (publiclaly available at https://www.epri.com/research/products/000000000001016737)
| |
| E9-13
| |
| | |
| License Amendment Request Enclosure 9 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Key Assumptions and Sources of Uncertainty
| |
| [5] QC-PRA-013, "Quad Cities Probabilistic Risk Assessment Summary Document Notebook,"
| |
| Revision 8, dated June 2019
| |
| [6] QC-MISC-047, "Assessment of Key Assumptions and Sources of Uncertainty for the Quad Cities Nuclear Power Plant," Revision 0
| |
| [7] NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,"
| |
| Volumes 1 and 2, dated September 2005 (ADAMS Accession Nos. ML15167A401 and ML15167A411)
| |
| [8] QC-PRA-021.62, "Quad Cities Nuclear Power Plant Fire PRA Uncertainty and Sensitivity Analysis Notebook," Revision 2, dated November 2021
| |
| [9] NSAC-161, "Faulted Systems Recovery Experience," dated May 1, 1992 (publically available at https://www.epri.com/research/products/NSAC-161)
| |
| E9-14
| |
| | |
| ENCLOSURE 10 License Amendment Request Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Program Implementation
| |
| | |
| License Amendment Request Enclosure 10 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Program Implementation
| |
| : 1. Introduction Section 4.0, Item 11 of the U.S. Nuclear Regulatory Commission (NRC) Final Safety Evaluation (Reference [1]) for NEI 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines" (Reference [2]),
| |
| requires that the license amendment request (LAR) discuss the programs and procedures for plant staff responsibilities for the RMTS implementation. Specifically, Item 11 requires a discussion of the decision process for Risk Management Action (RMA) implementation during a Risk Informed Completion Time (RICT).
| |
| This enclosure provides a description of the implementing programs and procedures regarding the plant staff responsibilities for the RICT Program, including training of plant personnel, and specifically discusses the decision process for RMA implementation during extended Completion Times (CTs).
| |
| : 2. RICT Program and Procedures Constellation Energy Generation, LLC (CEG) developed a fleet-level program description and implementing procedures for the RICT Program. The program description establishes the management responsibilities and general requirements for risk management, training, implementation, and monitoring of the RICT Program. More detailed procedures provide specific responsibilities, limitations, and instructions for implementing the RICT Program.
| |
| Procedures already implemented at other CEG sites will be used at QCNPS where applicable.
| |
| The program description and implementing procedures will incorporate the programmatic requirements for RMTS included in NEI 06-09. The program will be integrated with the online work control process. The work control process currently identifies the need to enter a LCO action statement as part of the planning process and will additionally identify whether the provisions of the RICT Program are required for the planned work. The risk thresholds associated with 10CFR50.65(a)(4) will be coordinated with the RICT limits. The maintenance rule performance monitoring provisions and Mitigating System Performance Index (MSPI) thresholds will assist in controlling the amount of risk expended in use of the RICT Program.
| |
| The Operations Department (licensed operators) is responsible for compliance with the Technical Specifications (TS) and will be responsible for implementation of RICTs and RMAs.
| |
| Entry into the RICT program will require management approval prior to pre-planned activities and as soon as practicable following emergent conditions.
| |
| The procedures for the RICT program address the following attributes consistent with NEI 06-09:
| |
| Plant management positions with authority to approve entry into the RICT Program Important definitions related to the RICT Program Departmental and position-specific responsibilities for activities in the RICT Program Plant conditions for which the RICT Program is applicable Limitations on implementing RICTs under voluntary and emergent conditions Implementation of the RICT Program 30-day back stop limit Use of the Real-Time Risk (RTR) tool E10-2
| |
| | |
| License Amendment Request Enclosure 10 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Program Implementation Guidance on recalculating RICT and Risk Management Action Time (RMAT) within 12 hours or within the most limiting front-stop CT after a plant configuration change Requirements to identify and implement RMAs when the RMAT is exceeded or is anticipated to be exceeded, and to consider common cause failure potential in emergent RICTs Guidance on the use of RMAs including the conditions under which they may be credited in RICT calculations Conditions for exiting a RICT Requirements for training on the RICT Program Documentation requirements related to individual RICT evaluations, implementation of extended CTs, and accumulated annual risk
| |
| : 3. RICT Program Training The scope of training for the RICT Program will include rules for the new TS program, RTR tool software, TS actions included in the program, and procedures. The following CEG personnel receive RICT Program training:
| |
| Site Personnel Operations Director Operations Personnel (Licensed and Non-Licensed)
| |
| Operations Training Outage Manager On-line Manager Planning and Scheduling Personnel Work Week Managers Regulatory Assurance Personnel Selected Maintenance Personnel Engineering Risk Management Other Selected Management Corporate Personnel Operations Corporate Functional Area Manager Fleet Outages Corporate Functional Area Manager Licensing Management and Personnel Risk Management Personnel and Managers Training Management and Personnel Other Selected Management Training is carried out in accordance with CEG training procedures and processes. These procedures were written based on the Institute of Nuclear Power Operations (INPO)
| |
| Accreditation (ACAD) requirements, as developed and maintained by the National Academy for Nuclear Training. CEG has three levels of training for implementation of the RICT Program as described below:
| |
| E10-3
| |
| | |
| License Amendment Request Enclosure 10 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Program Implementation Level 1 Training This is the most detailed training intended for the individuals who will be directly involved in the implementation of the RICT Program. This level of training includes the following attributes:
| |
| Specific training on the revised TS Record keeping requirements Case studies Hands-on experience with the Real-Time Risk tool for calculating RMAT and RICT Identifying appropriate RMAs Common cause failure RMA considerations in emergent RICTs Other detailed aspects of the RICT Program Level 2 Training This training is applicable to plant management positions with authority to approve entry into the RICT Program, as well as supervisors, managers, and other personnel who closely support RICT implementation. These individuals need a broad understanding of the purpose, concepts, and limitations of the RICT Program. Level 2 training is significantly more detailed than Level 3 training (described below), but it is different from Level 1 training in that hands-on time with the Real-Time Risk tool, case studies, and other specifics are not required.
| |
| Level 3 Training This training is intended for the remaining personnel who require an awareness of the RICT Program. These employees need basic knowledge of RICT Program requirements and procedures. This training will cover RICT Program concepts that are important to disseminate throughout the organization.
| |
| : 4. References
| |
| [1] Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines',"
| |
| dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| [2] Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)
| |
| Guidelines," Revision 0, dated October 12, 2012 (ADAMS Accession No. ML12286A322)
| |
| E10-4
| |
| | |
| ENCLOSURE 11 License Amendment Request Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Monitoring Program
| |
| | |
| License Amendment Request Enclosure 11 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Monitoring Program
| |
| : 1. Introduction Section 4.0, Item 12 of the U.S. Nuclear Regulatory Commission (NRC) Final Safety Evaluation (Reference [1]) for NEI 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines" (Reference [2]),
| |
| requires that the license amendment request (LAR) provide a description of the implementation and monitoring program as described in Regulatory Guide (RG) 1.174, Revision 1 (Reference [3]) and NEI 06-09 (Reference [2]). (Note that RG 1.174, Revision 3 (Reference [4]), issued by the NRC in January 2018, made editorial changes to the applicable section referenced in the NRC safety evaluation for Section 4.0, Item 12.)
| |
| This enclosure provides a description of the process applied to monitor the cumulative risk impact of implementation of the Risk Informed Completion Time (RICT) Program, specifically the calculation of cumulative risk of extended Completion Times (CTs). Step 14 of Section 2.3.1 and Step 7.1 of Section 2.3.2 of NEI 06-09 (Reference [2]) discusses calculation of the cumulative risk for the RICT Program. General requirements for a Performance Monitoring Program for risk-informed applications are discussed in Element 3 of RG 1.174 (Reference [3]).
| |
| : 2. Description of Monitoring Program The RICT Program will require calculation of cumulative risk impact at least every refueling cycle, not to exceed 24 months, consistent with the guidance in NEI 06-09 (Reference [2]).
| |
| Procedures and processes already implemented at other Constellation Energy Generation, LLC (CEG) sites will be used at Quad Cities Nuclear Power Station where applicable. For the assessment period under evaluation, data will be collected for the risk increase associated with each application of an extended CT for both core damage frequency (CDF) and large early release frequency (LERF), and the total risk will be calculated by summing all risk associated with each RICT application. This summation is the change in CDF or LERF above the zero maintenance baseline levels during the period of operation in the extended CT (i.e., beyond the front-stop CT). The change in risk will be converted to average annual values.
| |
| The total average annual change in risk for extended CTs will be compared to the guidance of RG 1.174, Figures 4 and 5 (Reference [4]), acceptance guidelines for CDF and LERF, respectively. If the actual annual risk increase is acceptable (i.e., not in Region I of Figures 4 and 5 of RG 1.174), then RICT program implementation is acceptable for the assessment period. Otherwise, further assessment of the cause of exceeding the acceptance guidelines of RG 1.174 and implementation of any necessary corrective actions to ensure future plant operation is within the guidelines will be conducted under the corrective action program.
| |
| The evaluation of cumulative risk will also identify areas for consideration, such as:
| |
| RICT applications that dominated the risk increase Risk contributions from planned vs. emergent RICT applications Risk Management Actions (RMAs) implemented but not credited in the risk calculations Risk impact from applying RICT to avoid multiple shorter duration outages Any specific RICT application that incurred a large proportion of the risk E11-2
| |
| | |
| License Amendment Request Enclosure 11 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Monitoring Program Based on a review of the considerations above, corrective actions will be developed and implemented as appropriate. These actions may include:
| |
| Administrative restrictions on the use of RICTs for specific high-risk configurations Additional RMAs for specific configurations Rescheduling planned maintenance activities Deferring planned maintenance to shutdown conditions Use of temporary equipment to replace out-of-service structures, systems, and components (SSCs)
| |
| Plant modifications to reduce risk impact of future planned maintenance configurations In addition to impacting cumulative risk, implementation of the RICT Program may potentially impact the unavailability of SSCs. The existing Maintenance Rule (MR) monitoring programs under 10 CFR 50.65(a)(1) and (a)(2) provide for evaluation and disposition of unavailability impacts which may be incurred from implementation of the RICT Program. The SSCs in the scope of the RICT Program which are also in the scope of the MR allows the use of the MR Program.
| |
| The monitoring program for the MR, along with the specific assessment of cumulative risk impact described above, serve as the Implementation and Monitoring Program for the RICT Program as described in Element 3 of RG 1.174 (Reference [3]) and NEI 06-09 (Reference [2]).
| |
| : 3. References
| |
| [1] Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines',"
| |
| dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| [2] Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)
| |
| Guidelines," Revision 0, dated October 12, 2012 (ADAMS Accession No. ML12286A322)
| |
| [3] Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, dated November 2002 (ADAMS Accession No. ML023240437)
| |
| [4] Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, dated January 2018 (ADAMS Accession No. ML17317A256)
| |
| E11-3
| |
| | |
| ENCLOSURE 12 License Amendment Request Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Risk Management Action Examples
| |
| | |
| License Amendment Request Enclosure 12 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Risk Management Action Examples
| |
| : 1. Introduction This enclosure describes the process for identification and implementation of Risk Management Actions (RMAs) applicable during extended Completion Times (CTs) and provides examples of RMAs. RMAs will be governed by plant procedures for planning and scheduling maintenance activities. The procedures will provide guidance for the determination and implementation of RMAs when entering the Risk Informed Completion Time (RICT) Program consistent with the guidance provided in Nuclear Energy Institute (NEI) 06-09, Revision 0-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)
| |
| Guidelines (Reference [1]).
| |
| : 2. Responsibilities For planned entries into the RICT Program, Work Management is responsible for developing the RMAs with assistance from Operations and Risk Management. Operations is responsible for approval and implementation of RMAs. For emergent entry into extended CTs, Operations is also responsible for developing the RMAs.
| |
| : 3. Procedural Guidance For planned maintenance activities, implementation of RMAs will be required if it is anticipated that the Risk Management Action Time (RMAT) will be exceeded. For emergent activities, RMAs must be implemented if the RMAT is reached. Also, if an emergent event occurs requiring recalculation of a RMAT already in place, the procedure will require a re-evaluation of the existing RMAs for the new plant configuration to determine if new RMAs are appropriate.
| |
| These requirements of the RICT Program are consistent with the guidance of NEI 06-09 [1].
| |
| For emergent entry into a RICT, if the extent of condition is not known, RMAs related to the success of redundant and diverse structures, systems, and components (SSCs) and reducing the likelihood of initiating events relying on the affected function will be developed to address the increased likelihood of a common cause event.
| |
| RMAs will be implemented in accordance with current Constellation Energy Generation, LLC (CEG) procedures (References [2], [3], [4], and [5]) no later than the time at which an incremental core damage probability (ICDP) of 1E-6 is reached, or no later than the time when an incremental large early release probability (ILERP) of 1E-7 is reached. If, as the result of an emergent condition, the instantaneous core damage frequency (ICDF) or the instantaneous large early release frequency (ILERF) exceeds 1E-3 per year or 1E-4 per year, respectively, RMAs are also required to be implemented. These requirements are consistent with the guidelines of NEI 06-09 (Reference [1]).
| |
| By determining which SSCs are most important from a core damage frequency (CDF) or large early release frequency (LERF) perspective for a specific plant configuration, RMAs may be created to protect these SSCs. Similarly, knowledge of the initiating event or sequence contribution to the configuration-specific CDF or LERF allows development of RMAs that enhance the capability to mitigate such events. The guidance in NUREG-1855, Revision 1, "Guidance on the treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking," (Reference [6]) and the Electric Power Research Institute (EPRI) topical E12-2
| |
| | |
| License Amendment Request Enclosure 12 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Risk Management Action Examples report (TR), TR-1026511, Revision 0, "Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty,"
| |
| (Reference [7]) will be used in examining Probabilistic Risk Assessment (PRA) results for significant contributors for the configuration, to aid in identifying appropriate compensatory measures (e.g., related to risk- significant systems that may provide diverse protection, or important support systems or human actions). Enclosure 9 identifies several areas of uncertainty in the internal events and fire PRAs that will be considered in defining configuration-specific RMAs when entering a RICT.
| |
| If the planned activity or emergent condition includes an SSC that is identified to impact fire PRA, as identified in the current Real-Time Risk Program, fire PRA specific RMAs associated with that SSC will be implemented per the current plant procedure.
| |
| It is possible to credit RMAs in RICT calculations, to the extent the associated plant equipment and operator actions are modeled in the PRA; however, such quantification of RMAs is neither required nor expected by NEI 06-09 (Reference [1]). Nonetheless, if RMAs will be credited to determine RICTs, the procedure instructions will be consistent with the guidance in NEI 06-09 (Reference [1]).
| |
| NEI 06-09 (Reference [1]) classifies RMAs into the three categories described below:
| |
| : 1) Actions to increase risk awareness and control.
| |
| Shift brief Pre-job brief Training Presence of system engineer or other expertise related to the activity Special purpose procedure to identify risk sources and contingency plans
| |
| : 2) Actions to reduce the duration of maintenance activities.
| |
| Pre-staging materials Conducting training on mock-ups Performing the activity around the clock Performing walk-downs on the actual system(s) to be worked on prior to beginning work
| |
| : 3) Actions to minimize the magnitude of the risk increase.
| |
| Suspend or minimize activities on redundant systems Suspend or minimize activities on other systems that adversely affect the CDF or LERF Suspend or minimize activities on systems that may cause a trip or transient to minimize the likelihood of an initiating event that the out-of-service component is meant to mitigate Use temporary equipment to provide backup power, ventilation, etc.
| |
| Reschedule other risk-significant activities E12-3
| |
| | |
| License Amendment Request Enclosure 12 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Risk Management Action Examples Determination of RMAs involves the use of both qualitative and quantitative considerations for the specific plant configuration and the practical means available to manage risk. The scope and number of RMAs developed and implemented are reached in a graded manner.
| |
| Procedural guidance for development of RMAs in support of the RICT program builds off the RMAs developed for other processes, such as the RMAs developed under the 10 CFR 50.65(a)(4) program and the protected equipment program. Additionally, Common Cause RMAs are developed to address the potential impact of common cause failures.
| |
| General RMAs are developed for input into the RICT system guidelines. These guidelines are listed in site-specific Training & Reference Materials (T&RM) and are developed using a graded approach. Consideration is given for system functionality and includes consideration for common cause impacts within the system. These RMAs include:
| |
| Consideration of rescheduling maintenance to reduce risk Discussion of RICT in pre-job briefs Consideration of proactive return-to-service of other equipment Efficient execution of maintenance In addition to the RMAs developed qualitatively for the system guidelines, RMAs are developed based on the Real-Time Risk (RTR) tool to identify configuration-specific RMA candidates to manage the risk associated with internal events, internal flooding, and fire events. These actions include:
| |
| Identification of important equipment or trains for protection Identification of important operator actions for briefings Identification of key fire initiators and fire zones for RMAs in accordance with the site Fire RMA process Identification of dominant initiating events and actions to minimize potential for initiators Consideration of insights from PRA model cutsets, through comparison of importances Common cause RMAs are also developed to ensure availability of redundant SSCs, to ensure availability of diverse or alternate systems, to reduce the likelihood of initiating events that require operation of the out-of-service components, and to prepare plant personnel to respond to additional failures. Common cause RMAs are developed by considering the impact of loss of function for the affected SSCs.
| |
| Examples of common cause RMAs include:
| |
| Performance of non-intrusive inspections on alternate trains Confidence runs performed for standby SSCs Increased monitoring for running components Expansion of monitoring for running components Deferring maintenance and testing activities that could generate an initiating event which would require operation of potentially affected SSCs Readiness of operators and maintenance to respond to additional failures E12-4
| |
| | |
| License Amendment Request Enclosure 12 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Risk Management Action Examples Shift briefs or standing orders which focus on initiating event response or loss of potentially affected SSCs Per CEG procedures, for emergent conditions where the extent of condition is not performed prior to entering into the RMATs or the extent of condition cannot rule out the potential for common cause failure, common cause RMAs are expected to be implemented to mitigate common cause failure potential and impact. These can include the pre-identified RMAs included in the system guidelines as discussed above, as well as alternative common cause RMAs for the specific configuration. Alternate RMAs, including both regular and common cause considerations, are developed for the specific configuration following the steps outlined above.
| |
| : 4. Examples Representative examples of RMAs that may be considered during a RICT Program entry to reduce the risk impact and ensure adequate defense-in-depth are provided below. Specific examples are given for electrical equipment and one Low Pressure Coolant Injection (LPCI) pump.
| |
| 4.1 Electrical Action Statements 4.1.1 For TS 3.8.1.A, one required offsite circuit inoperable, additional RMAs would include:
| |
| : 1. Actions to increase risk awareness and control.
| |
| Briefing of the on-shift Operations crew concerning the unit activities, including any compensatory measures established, and review of the appropriate emergency operating procedures for a Loss of Offsite Power (LOOP) and station blackout (SBO) including bus crossties.
| |
| Notification of the transmission system operator (TSO) of the configuration so that any planned activities with the potential to cause a grid disturbance are deferred.
| |
| Proactive implementation of RMAs during times of high grid stress conditions prior to reaching the RMAT, such as during high demand conditions.
| |
| : 2. Actions to reduce the duration of maintenance activities.
| |
| For preplanned RICT entry, creation of a fragnet1 which is reviewed for personnel resource availability.
| |
| Confirmation of parts availability prior to entry into a preplanned RICT.
| |
| Walkdown of work prior to execution.
| |
| : 3. Actions to minimize the magnitude of the risk increase.
| |
| Evaluation of weather conditions for threats to the reliability of remaining offsite power supplies.
| |
| 1 A sub-tiered schedule of detailed activities required to accomplish the desired tasks.
| |
| E12-5
| |
| | |
| License Amendment Request Enclosure 12 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Risk Management Action Examples Deferral of elective maintenance in the switchyard, on the station electrical distribution systems, and on the main and auxiliary transformers associated with the unit.
| |
| Protection of the remaining offsite source, including switchyard and transformer.
| |
| Deferral of planned maintenance or testing that affects the reliability of DGs and their associated support equipment. Treat these as protected equipment.
| |
| Implementation of 10 CFR 50.65(a)(4) fire-specific RMAs associated with the affected offsite source.
| |
| 4.1.2 For TS 3.8.1.B, one required DG inoperable, additional RMAs would include:
| |
| : 1. Actions to increase risk awareness and control.
| |
| Briefing of the on-shift Operations crew concerning the unit activities, including any compensatory measures established, and review of the appropriate emergency operating procedures for a LOOP and SBO including bus crossties.
| |
| Notification of the TSO of the configuration so that any planned activities with the potential to cause a grid disturbance are deferred.
| |
| Proactive implementation of RMAs during times of high grid stress conditions prior to reaching the RMAT, such as during high demand conditions.
| |
| : 2. Actions to reduce the duration of maintenance activities.
| |
| For preplanned RICT entry, creation of a fragnet which is reviewed for personnel resource availability.
| |
| Confirmation of parts availability prior to entry into a preplanned RICT.
| |
| Walkdown of work prior to execution.
| |
| : 3. Actions to minimize the magnitude of the risk increase.
| |
| Evaluation of weather conditions for threats to the reliability of remaining offsite power supplies.
| |
| Deferral of elective maintenance in the switchyard, on the station electrical distribution systems, and on the main and auxiliary transformers associated with the unit.
| |
| Deferral of planned maintenance or testing that affects the reliability of operable DGs and their associated support equipment. Treat the remaining operable DGs as protected equipment.
| |
| Deferral of planned maintenance or testing on redundant train safety systems. If testing or maintenance activities must be performed, a review of the potential risk impact will be performed.
| |
| Implementation of 10 CFR 50.65(a)(4) fire-specific RMAs associated with the affected DG.
| |
| E12-6
| |
| | |
| License Amendment Request Enclosure 12 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Risk Management Action Examples 4.1.3 For TS 3.8.1.D, one required offsite circuit inoperable AND one required DG inoperable, additional RMAs would include:
| |
| : 1. Actions to increase risk awareness and control.
| |
| Briefing of on-shift Operations crew concerning the unit activities, including any compensatory measures established, and review of the appropriate abnormal operating procedures for a LOOP and SBO.
| |
| Notification of the TSO of the configuration so that any planned activities with the potential to cause a grid disturbance are deferred.
| |
| Proactive implementation of RMAs during times of high grid stress conditions prior to reaching the RMAT, such as during high demand conditions.
| |
| For a planned RICT, prior to removal from service, the actions in the associated loss of bus procedure with the inoperable DG would be reviewed.
| |
| : 2. Actions to reduce the duration of maintenance activities.
| |
| For preplanned RICT entry, creation of a fragnet which is reviewed for personnel resource availability.
| |
| Confirmation of parts availability prior to entry into a preplanned RICT.
| |
| Walkdown of work prior to execution.
| |
| : 3. Actions to minimize the magnitude of the risk increase.
| |
| Deferral of elective maintenance in the switchyard, on the station electrical distribution systems, and on the main and station reserve transformers associated with the unit.
| |
| Deferral of planned maintenance or testing that affects the reliability of DGs and their associated support equipment for the remaining buses.
| |
| Implementation of 10 CFR 50.65(a)(4) fire-specific RMAs associated with the affected DG.
| |
| Place unaffected trains of systems into service. For example, if one of two Control Rod Drive Hydraulic (CRD) pumps is powered by the affected bus, the other unaffected pump would be placed in service to support control rod manipulation. This would be done prior to entry into a planned RICT.
| |
| 4.1.4 TS 3.8.4.D, Division 1 or 2 125 VDC electrical power subsystem inoperable for reasons other than Conditions B or C, additional RMAs would include:
| |
| : 1. Actions to increase risk awareness and control.
| |
| Briefing of on-shift Operations crew concerning the unit activities, including any compensatory measures established, and review of the appropriate emergency operating procedures for a Loss of DC division and SBO.
| |
| Briefing of on-shift Operations crew concerning the impact the DC division has on potential response to plant events, such as reduced control systems.
| |
| E12-7
| |
| | |
| License Amendment Request Enclosure 12 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Risk Management Action Examples For a planned RICT, prior to removal from service, the actions in the associated loss of bus procedure would be reviewed.
| |
| Minimize activities that could trip the unit.
| |
| : 2. Actions to reduce the duration of maintenance activities.
| |
| For preplanned RICT entry, creation of a fragnet which is reviewed for personnel resource availability.
| |
| Confirmation of parts availability prior to entry into a preplanned RICT.
| |
| Walkdown of work prior to execution.
| |
| : 3. Actions to minimize the magnitude of the risk increase.
| |
| Deferral of elective maintenance in the switchyard, on the station electrical distribution systems, and on the main and station reserve transformers associated with the unit.
| |
| Protection of the remaining DC electrical buses.
| |
| Remove nonessential loads from battery to extend time voltage to remain above minimum required level.
| |
| Implementation of 10 CFR 50.65(a)(4) fire-specific RMAs associated with the affected bus.
| |
| 4.2 Emergency Core Cooling Systems (ECCS) Action Statements 4.2.1 For TS 3.5.1.C, one Low Pressure Coolant Injection (LPCI) pump in each subsystem inoperable, additional RMAs would include:
| |
| Defer planned maintenance or testing activities on the redundant LPCI pumps and associated support equipment. Treat those systems as protected equipment.
| |
| Defer planned maintenance or testing that affects the reliability of those safety systems that provide defense-in-depth. If testing or maintenance activities must occur, a review of the potential risk impact will be performed.
| |
| Minimize activities that could trip the unit.
| |
| Verify system alignment of remaining LPCI pumps and LPCI subsystems.
| |
| Implement 10 CFR 50.65(a)(4) fire-specific RMAs associated with the LPCI subsystems.
| |
| : 5. References
| |
| [1] NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," Revision 0-A, dated October 12, 2012 (ADAMS Accession No. ML12286A322)
| |
| [2] OP-AA-201-012-1001, "Operations On-Line Fire Risk Management," Revision 5, dated December 2022.
| |
| [3] OP-AA-108-118, "Risk Informed Completion Time," Revision 1, dated April 2020 E12-8
| |
| | |
| License Amendment Request Enclosure 12 Adopt Risk Informed Completion Times TSTF-505 Docket Nos. 50-254 and 50-265 Risk Management Action Examples
| |
| [4] OP-AA-108-117, "Protected Equipment Program," Revision 7, dated February 2021
| |
| [5] WC-AA-101-1006, "On-Line Risk Management and Assessment," Revision 4, dated May 2020
| |
| [6] NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," U.S. Nuclear Regulatory Commission, Revision 1, dated March 2017 (ADAMS Accession No. ML17062A466)
| |
| [7] EPRI TR-1026511, "Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty," dated December 2012 (publicly available at https://www.epri.com/research/products/000000000001026511)
| |
| E12-9}}
| |
