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| {{#Wiki_filter::) entergy Phil Couture Senior Manager Regulatory Assurance 601-368-5102 GNRO-2023/00014 10 CFR 50.90 June 6, 2023 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 | | {{#Wiki_filter:}} |
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| ==Subject:==
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| 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" Grand Gulf Nuclear Station, Unit 1 NRC Docket No. 50-416 Renewed License No. NPF-29 Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy or EOI) is requesting an amendment to the Technical Specifications (TS) of the Grand Gulf Nuclear Station Unit 1 (GGNS).
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| The proposed amendment would modify TS requirements to permit the use of Risk Informed Completion Times in accordance with TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b (ADAMS Accession No. ML18183A493). A model safety evaluation with Limitations and Conditions was provided by the Nuclear Regulatory Commission (NRC) to the TS Task Force (TSTF) on November 21, 2018 (ADAMS Accession No. ML18253A085).
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| * Attachment 1 provides a description and assessment of the proposed change, the requested confirmation of applicability, and plant-specific verifications.
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| * Attachment 2 provides the existing TS pages marked up to show the proposed changes.
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| * Attachment 3 provides existing TS Bases pages marked up to show the proposed changes and is provided for information only.
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| * Attachment 4 provides a cross-reference between the TS included in TSTF-505, Revision 2, and the GGNS plant-specific TS.
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| * Attachment 5 provides information supporting the redundant means available to mitigate accidents for instrumentation governed by the TS that are proposed to be included as part of the RICT program in this submittal.
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| * Attachment 6 provides a list of implementation items that must be completed prior to implementing the Risk-Informed Completion Time Program at GGNS.
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| GNRO-2023/00014 Page 2 of 3 Entergy requests approval of the proposed license amendment by July 15, 2024, with the amendment being implemented within 180 days of issuance of amendment.
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| In accordance with 10 CFR 50.91, Entergy is notifying the State of Mississippi of this amendment request by transmitting a copy of this letter and attachments and enclosures to the designated State Official.
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| This letter and its attachments and enclosures contain no regulatory commitments.
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| If there are any questions regarding this submittal, please contact me at (601) 368-5102.
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| I declare under penalty of perjury that the foregoing is true and correct.
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| Executed on June 6, 2023.
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| Respectfully, Philip Digitally signed by Philip Couture Couture Date: 2023.06.06 15:13:41 -05'00' Phil Couture PC/ram Attachments: 1. Description and Assessment
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| : 2. Technical Specification Page Markups
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| : 3. Technical Specification Bases Page Markups - For Information Only
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| : 4. Cross-Reference of TSTF-505 and GGNS Technical Specifications
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| : 5. Evaluation of Instrumentation and Control Systems
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| : 6. GGNS RICT Program PRA Implementation Items
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| ==Enclosures:==
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| : 1. List of Revised Required Actions to Corresponding PRA Functions
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| : 2. Information Supporting Consistency with Regulatory Guide 1.200, Revision 2
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| : 3. Information Supporting Technical Adequacy of PRA Models Without PRA Standards Endorsed by Regulatory Guide 1.200, Revision 2
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| : 4. Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
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| : 5. Baseline CDF and LERF
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| : 6. Justification of Application of At-Power PRA Models to Shutdown Modes
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| : 7. PRA Model Update Process
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| : 8. Attributes of the Real-Time Model
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| : 9. Key Assumptions and Sources of Uncertainty
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| : 10. Program Implementation
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| : 11. Monitoring Program
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| : 12. Risk Management Action Examples
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| GNRO-2023/00014 Page 3 of 3 cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector - Grand Gulf Nuclear Station NRC Project Manager - Grand Gulf Nuclear Station State Health Officer, Mississippi Department of Health
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| Attachment 1 GNRO-2023/00014 Description and Assessment (10 pages follow)
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| GNRO-2023/00014 Page 1 of 10 TABLE OF CONTENTS DESCRIPTION AND ASSESSMENT .......................................................................................... 2 1.0
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| ==SUMMARY==
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| DESCRIPTION ................................................................................................ 2 2.0 ASSESSMENT ................................................................................................................... 2 2.1 APPLICABILITY OF PUBLISHED SAFETY EVALUATION........................................................ 2 2.2 VERIFICATIONS AND REGULATORY COMMITMENTS.......................................................... 2 2.3 OPTIONAL VARIATIONS.................................................................................................. 3 3.0 REGULATORY SAFETY ANALYSIS .................................................................................. 7 3.1 NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS ................................................... 7
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| ==3.2 CONCLUSION==
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| S.............................................................................................................. 9
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| ==4.0 ENVIRONMENTAL CONSIDERATION==
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| ............................................................................... 9
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| ==5.0 REFERENCES==
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| ................................................................................................................. 10
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| GNRO-2023/00014 Page 2 of 10 DESCRIPTION AND ASSESSMENT 1.0
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| ==SUMMARY==
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| DESCRIPTION The proposed amendment would modify the Technical Specification (TS) requirements related to Completion Times (CTs) for Required Actions to provide the option to calculate a longer, risk-informed CT (RICT). A new program, the Risk-Informed Completion Time Program, is added to TS Section 5.0, Administrative Controls.
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| 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)
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| Guidelines," Revision 0, which was approved by the NRC on May 17, 2007 (Reference 1).
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| Adherence to NEI 06-09-A is required by the RICT Program.
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| The proposed amendment is consistent with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" (Reference 2). However, only those Required Actions described in Attachment 4 and Enclosure 1, as reflected in the proposed TS mark-ups provided in Attachments 2 and 3, are proposed to be changed. This is because some of the modified Required Actions in TSTF-505 are not applicable to Grand Gulf Nuclear Station Unit 1 (GGNS), and there are some plant-specific Required Actions not included in TSTF-505 that are included in this proposed amendment.
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| 2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation Entergy Operations, Inc. (Entergy or EOI) has reviewed TSTF-505, Revision 2, and the model safety evaluation dated November 21, 2018 (Reference 3). This review included the supporting information provided to support TSTF-505 and the safety evaluation for NEI 06-09-A. As described in the subsequent paragraphs, Entergy has concluded that the technical basis is applicable to GGNS and supports incorporation of this amendment in the GGNS TS.
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| 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:
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| : 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.
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| : 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 discussed in Regulatory Guide (RG) 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, Section 4.2.
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| GNRO-2023/00014 Page 3 of 10
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| : 3. Enclosure 3 is not applicable since each PRA model used for the RICT Program is addressed using a standard endorsed by the NRC.
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| : 4. Enclosure 4 provides appropriate justification for excluding sources of risk not addressed by the PRA models.
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| : 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.
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| : 6. Enclosure 6 is not applicable since the RICT Program is not being applied to shutdown modes.
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| : 7. Enclosure 7 provides a discussion of Entergys programs and procedures that will assure the PRA models that support the RICT Program are maintained consistent with the as-built, as-operated plant.
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| : 8. Enclosure 8 provides a description of how the baseline PRA model, which calculates average annual risk, is evaluated and modified to assess real-time configuration risk, and describes the scope of, and quality controls applied to the real-time model.
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| : 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.
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| : 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.
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| : 11. Enclosure 11 provides a description of the implementation and monitoring program as described in NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, Section 2.3.2, Step 7.
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| : 12. Enclosure 12 provides a description of the process to identify and provide RMAs.
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| 2.3 Optional Variations Entergy is proposing the following variations from the TS changes described in TSTF-505, Revision 2, or the applicable parts of the NRC staffs model safety evaluation dated November 21, 2018. These options were recognized as acceptable variations in TSTF-505 and the NRC staffs model safety evaluation.
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| Note that, in a few instances, the GGNS TS utilizes different numbering and titles than the Standard Technical Specifications on which TSTF-505 was based. These differences are administrative and do not affect the applicability of TSTF-505 to the GGNS TS. Attachment 4 provides specific information.
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| GGNS is a BWR/6 design, its Technical Specifications align with NUREG-1434, Standard Technical Specifications, General Electric BWR/6 Plants (Reference 4). Attachment 4 is a
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| GNRO-2023/00014 Page 4 of 10 cross-reference that provides a comparison between NUREG-1434 and the GGNS Technical Specifications 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 Actions. The cross-reference in Attachment 4 identifies the following:
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| : 1. GGNS Required Actions that have identical numbers to the corresponding NUREG-1434 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 NRCs model safety evaluation dated November 21, 2018.
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| : 2. GGNS Required Actions that have different numbering than the NUREG-1434 Required Actions are an administrative deviation from TSTF-505 with no impact on the NRCs model safety evaluation dated November 21, 2018.
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| : 3. For NUREG-1434 Required Actions that are not contained in the GGNS TS, the corresponding TSTF-505 mark-ups for the Required Actions are not applicable to GGNS. This is an administrative deviation from TSTF-505 with no impact on the NRCs model safety evaluation dated November 21, 2018.
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| : 4. The model application provided in TSTF-505 includes an attachment for clean typed TS pages reflecting the proposed changes. GGNS 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 NRCs model safety evaluation dated November 21, 2018.
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| Because of this deviation, the contents and numbering of the attachments for this amendment request differ from the attachments specified in the model application in TSTF-505.
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| : 5. As the proposed GGNS RICT Program is applicable in Modes 1 and 2, GGNS will not adopt changes in TSTF-505 for Required Actions that are only applicable in Modes 3 and below.
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| : 6. The model application provided in TSTF-505 includes mark-ups to Completion Times for NUREG-1434 in a format using an OR Logical Connector followed by In accordance with the Risk Informed Completion Time Program. Several existing Required Actions have two Completion Times connected by the Logical Connector AND in the current GGNS TS. GGNS TS Section 1.2, Logical Connectors, specifies that Completion Times only use first level logic. Therefore, the proposed markups have been modified for these Required Actions to embed or in accordance with the Risk Informed Completion Time Program into the existing Completion Times. This follows GGNS TS Section 1.2 and does not create a second level logic for the Completion Times. This is an administrative deviation from TSTF-505 with no impact on the NRCs model safety evaluation dated November 21, 2018. This administrative deviation is consistent with TS Amendments issued for Exelons LaSalle County Station (ML21162A069), Clinton Power
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| GNRO-2023/00014 Page 5 of 10 Station (ML21132A288), and Northern States Power Companys Monticello Generating Plant (ML21148A274).
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| : 7. For one TS, as noted in the Attachment 2 TS markups, a Note 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 safety 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 system, GGNS has chosen to replace the TSTF-505 Note with a Note which reads Not applicable when a loss of function occurs, which accomplishes the intended purpose of the TSTF-505 Note.
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| : 8. GGNS proposes to make two minor editorial changes to the Required Action in the new proposed Example 1.3-8 relative to the Example 1.3-8 provided in TSTF-505, Revision
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| : 2. TSTF-505, Revision 2, Example 1.3-8 provides a standard default Condition of Required Action and associated Completion Time not met, with a Required Action of Be in MODE 3 [in 6 hours] AND Be in MODE 5 [in 36 hours]. However, the typical default Required Action found in TS Section 1.3 Examples as well as TS Section 3 Limiting Conditions for Operation (LCOs) is Be in MODE 3 [in 12 hours] AND Be in MODE 4 [in 36 hours]. Therefore, as noted in the TS markups in Attachment 2, GGNS proposes to change the Completion Time for the MODE 3 Required Action to 12 hours and to specify MODE 4 rather than MODE 5 in the Required Action of the default Condition in the new proposed Example 1.3-8. These changes are administrative in nature and do not affect the applicability of TSTF-505, Revision 2, to the GGNS TS.
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| : 9. TSTF-505, Revision 2, insert for the proposed RICT Program, Item e describes requirements regarding risk assessment approaches and methods as it relates to this application. Within this item is the statement that Methods to assess the risk from extending the Completion Times must be PRA methods used to support this license amendment, or other methods . Entergy considers the phrase used to support this license amendment to be potentially confusing since there is no indication as to which license amendment is being referred to in this paragraph. Therefore, for clarification purposes, Entergy proposes to modify this statement to read Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program in Amendment No. [###], or other methods .
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| : 10. There are several plant-specific LCOs and associated Required Actions for which GGNS is proposing to apply the RICT Program that are variations from TSTF-505, Revision 2.
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| These TS Actions are identified in Attachment 4 with additional justification provided below.
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| * 3.6.1.6 - Low-Low Set (LLS) Valves LCO: The LLS function of six safety/relief valves shall be OPERABLE.
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| Condition A: One LLS valve inoperable.
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| TSTF-505 does not apply a RICT to TS 3.6.1.6. GGNS proposes to apply a RICT to Condition A of this TS because the safety/relief valves are modeled in the PRA and the
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| GNRO-2023/00014 Page 6 of 10 LLS function can be conservatively modeled in the Configuration Risk Management Program (CRMP) model. There is precedent for including TS 3.6.1.6, Required Action A.1 in the RICT program; the NRC approved application of a RICT to TS LCO 3.6.1.6 Condition A for Exelons Clinton Power Station, Unit 1, in Amendment 238 dated June 28, 2021 (ML21132A288).
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| Required Action A.1 requires the inoperable LLS valve be restored to OPERABLE status within a 14-day Completion Time. With one LLS valve inoperable, the remaining OPERABLE LLS valves are adequate to perform the design function described below.
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| Six of the safety/relief valves (S/RVs) are equipped to provide the LLS function. The LLS logic causes two LLS valves to be opened at a lower pressure than the relief or safety mode pressure setpoints and causes all LLS valves to stay open longer, such that reopening of more than one S/RV is prevented on subsequent actuations. The LLS function prevents excessive short duration S/RV cycles with valve actuation at the relief setpoint. As indicated in Table E1-1 of Enclosure 1, the configurations associated with TS 3.6.1.6 Condition A will be conservatively modeled for calculating a RICT. Required Action A.1 restores the inoperable LLS valves to OPERABLE status; therefore, this Action meets the requirements for inclusion in the RICT Program as outlined in TSTF-505, Revision 2.
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| * TS 3.8.1 - AC Sources - Operating LCO: The following AC electrical power sources shall be OPERABLE:
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| : a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electric Power Distribution System;
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| : b. Three diesel generators (DGs); and
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| : c. Division 1 and Division 2 automatic load sequencers.
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| Condition B: One required DG inoperable for reasons other than Condition F.
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| Required Action B.4: Restore required DG to OPERABLE status.
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| Completion Time B.4: 72 hours from discovery of an inoperable Division 3 DG AND 14 days For the Condition of one required DG inoperable, NUREG-1434 allows a Completion Time of 72 hours to restore the required DG to OPERABLE status (LCO 3.8.1 Required Action B.4 of NUREG-1434). GGNS Completion Time B.4, which allows 72 hours from discovery of an inoperable Division 3 DG AND 14 days to restore the required DG to OPERABLE status, is a plant specific Completion Time not included in NUREG-1434 and therefore, not in TSTF-505.
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| This Completion Time was implemented as a result of Amendment 151 (Reference 5) dated July 16, 2002, which extended the Completion Time for an inoperable DG to provide needed flexibility in performing both corrective and preventative maintenance on the Division 1 and 2 DGs during power operation. The Completion Time to restore one required inoperable DG to OPERABLE status was revised to: 72 hours from discovery
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| GNRO-2023/00014 Page 7 of 10 of an inoperable Division 3 DG AND 14 days AND 17 days from discovery of failure to meet LCO.
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| Amendment 226 (Reference 6) removed the AND 17 days from discovery of failure to meet LCO from this TS Completion Time. Amendment 226 adopted TSTF-439, Revision 2, Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO, (Reference 7) into the GGNS TS. Adoption of TSTF-439 is a prerequisite to implementation of TSTF-505.
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| Therefore, this variation to TSTF-505, Revision 2, is justified through the implementation of Amendments 151 and 226.
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| Entergy has determined that the application of a RICT for these GGNS plant-specific TS Actions is consistent with TSTF-505, Revision 2, and with the NRCs model safety evaluation dated November 21, 2018. Application of a RICT for these plant-specific TS Actions will be controlled under the proposed RICT Program. The RICT Program provides the necessary administrative controls to permit extension of CTs 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 shut down the reactor, are also unchanged; only the TS Action CTs may be extended within the governance of the RICT Program.
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| 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, Revision 0 which was approved by the NRC on May 17, 2007. The NEI 06-09-A, Revision 0, 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 RMAs to maintain the risk below acceptable regulatory risk thresholds.
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| Therefore, the proposed application of a RICT to the above GGNS plant-specific TS Actions is consistent with TSTF-505, Revision 2 and with the NRC staffs model safety evaluation dated November 21, 2018.
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| Entergy has reviewed these proposed changes and determined that they do not affect the applicability of TSTF-505, Revision 2 to the GGNS TS.
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| 3.0 REGULATORY SAFETY ANALYSIS 3.1 No Significant Hazards Consideration Analysis Entergy Operations, Inc. (Entergy or EOI) has evaluated the proposed change to the TS using the criteria in 10 CFR 50.92 and has determined that the proposed change does not involve a significant hazards consideration.
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| Grand Gulf Nuclear Station Unit 1 (GGNS) requests adoption of an approved change to the standard technical specifications (STS) and plant-specific technical specifications (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
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| GNRO-2023/00014 Page 8 of 10 new program in Chapter 5, Administrative Controls, entitled the Risk-Informed Completion Time Program.
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| As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:
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| : 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
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| Response: No.
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| The proposed change permits 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 change does not involve a significant increase in the probability of an accident previously evaluated because the change involves no change to the plant or its modes of operation. The proposed change does 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.
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| Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
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| : 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
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| Response: No.
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| The proposed changes do not change the design, configuration, or method of operation of the plant. The proposed changes do not involve a physical alteration of the plant (no new or different kind of equipment will be installed.)
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| Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
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| : 3. Do the proposed changes involve a significant reduction in a margin of safety?
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| Response: No.
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| 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
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| GNRO-2023/00014 Page 9 of 10 management program considers cumulative effect of multiple systems or components being out of service and does so more effectively than the current TS.
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| Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
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| 1.0 Based on the above, Entergy concludes that the proposed changes present no significant hazards consideration under standards set forth in 10 CFR 50.92©, and, accordingly, a finding f "no significant hazards consideraton" is justifie.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.
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| ==4.0 ENVIRONMENTAL CONSIDERATION==
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| Entergy has reviewed the environmental evaluation included in the model safety evaluation published on November 21, 2018 (ADAMS Accession Mo. ML18267A259) as part of the Notice of Availability. Entergy has concluded that the NRC staff findings presented in that evaluation are applicable to GGNS.
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| 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 change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
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| Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
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| GNRO-2023/00014 Page 10 of 10
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| ==5.0 REFERENCES==
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| : 1. NRC letter to 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) Guideline, (ADAMS Accession No. ML071200238), dated May 17, 2007.
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| : 2. TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b (ADAMS Accession No. ML18183A493).
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| : 3. NRC letter to Technical Specifications Task Force, "Final Revised Model Safety Evaluation of Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," (ADAMS Accession No. ML18253A085),
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| dated November 21, 2018.
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| : 4. NRC NUREG-1434, "Standard Technical Specifications, General Electric BWR/6 Plants," Revision 5, Volumes 1 and 2, dated September 2021
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| : 5. Letter from Jaffe (NRC) to Eaton (Entergy), "Grand Gulf Nuclear Station, Unit 1 -
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| Issuance of Amendment Re: Extended Allowed Outage Time for Diesel Generators,"
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| (ADAMS Accession No. ML021860203), dated July 16, 2002.
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| : 6. Letter from Lingham (NRC) to Halter (Entergy), "Grand Gulf Nuclear Station, Unit 1 and River Bend Station, Unit 1 - Issuance of Amendments Re: Adoption of TSTF-439, Revision 2, 'Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO,'" (ADAMS Accession No. ML21011A068), dated February 2, 2021.
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| : 7. TSTF-439, Revision 2, "Eliminate Second Completion Times Limiting Time From Discovery of Failure to Meet an LCO," (ADAMS Accession No. ML051860296), dated June 20, 2005.
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| Attachment 2 GNRO-2023/00014 (Technical Specification Page Markups)
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| (43 pages follow)
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| INSERT EXAMPLE 1.3-8 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 met. B.2 Be in MODE 4. 36 hours 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.
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| 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 B must also be entered.
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| 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.
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| 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.
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| 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.
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| INSERT RICT 1 OR In accordance with the Risk Informed Completion Time Program INSERT RICT 2 or in accordance with the Risk Informed Completion Time Program INSERT RICT 3 OR
| |
| -------NOTE-------
| |
| Not applicable when a loss of function occurs.
| |
| In accordance with the Risk Informed Completion Time Program
| |
| | |
| -27)686-'8463+6%1
| |
| 6MWO-RJSVQIH'SQTPIXMSR8MQI4VSKVEQ 8LMWTVSKVEQTVSZMHIWGSRXVSPWXSGEPGYPEXIE6MWO-RJSVQIH'SQTPIXMSR8MQI 6-'8 ERHQYWXFIMQTPIQIRXIHMREGGSVHERGI[MXL2)- %
| |
| 6IZMWMSR 6MWO1EREKIH8IGLRMGEP7TIGMJMGEXMSRW 6187 +YMHIPMRIW8LI TVSKVEQWLEPPMRGPYHIXLIJSPPS[MRK
| |
| E 8LI6-'8QE]RSXI\GIIHHE]W
| |
| F %6-'8QE]SRP]FIYXMPM^IHMR13()7ERH
| |
| G ;LIRE6-'8MWFIMRKYWIHER]GLERKIXSXLITPERXGSRJMKYVEXMSREW HIJMRIHMR2)- %%TTIRHM\%QYWXFIGSRWMHIVIHJSVXLIIJJIGX SRXLI6-'8
| |
| *SVTPERRIHGLERKIWXLIVIZMWIH6-'8QYWXFIHIXIVQMRIHTVMSV XSMQTPIQIRXEXMSRSJXLIGLERKIMRGSRJMKYVEXMSR
| |
| *SVIQIVKIRXGSRHMXMSRWXLIVIZMWIH6-'8QYWXFIHIXIVQMRIH
| |
| [MXLMRXLIXMQIPMQMXWSJXLI6IUYMVIH%GXMSR'SQTPIXMSR8MQI MI
| |
| RSXXLI6-'8 SVLSYVWEJXIVXLITPERXGSRJMKYVEXMSRGLERKI
| |
| [LMGLIZIVMWPIWW
| |
| 6IZMWMRKXLI6-'8MWRSXVIUYMVIHMJXLITPERXGSRJMKYVEXMSRGLERKI
| |
| [SYPHPS[IVTPERXVMWOERH[SYPHVIWYPXMREPSRKIV6-'8
| |
| H *SVIQIVKIRXGSRHMXMSRWMJXLII\XIRXSJGSRHMXMSRIZEPYEXMSRJSV MRSTIVEFPIWXVYGXYVIWW]WXIQWSVGSQTSRIRXW 77'W MWRSXGSQTPIXI TVMSVXSI\GIIHMRKXLI'SQTPIXMSR8MQIXLI6-'8WLEPPEGGSYRXJSVXLI MRGVIEWIHTSWWMFMPMX]SJGSQQSRGEYWIJEMPYVI ''* F]IMXLIV
| |
| 2YQIVMGEPP]EGGSYRXMRKJSVXLIMRGVIEWIHTSWWMFMPMX]SJ''*MRXLI 6-'8GEPGYPEXMSR SV
| |
| 6MWO1EREKIQIRX%GXMSRW 61%W RSXEPVIEH]GVIHMXIHMRXLI 6-'8GEPGYPEXMSRWLEPPFIMQTPIQIRXIHXLEXWYTTSVXVIHYRHERXSV HMZIVWI77'WXLEXTIVJSVQXLIJYRGXMSR W SJXLIMRSTIVEFPI77'W
| |
| ERHMJTVEGXMGEFPIVIHYGIXLIJVIUYIRG]SJMRMXMEXMRKIZIRXWXLEX GLEPPIRKIXLIJYRGXMSRW W TIVJSVQIHF]XLIMRSTIVEFPI77'W
| |
| I 8LIVMWOEWWIWWQIRXETTVSEGLIWERHQIXLSHWWLEPPFIEGGITXEFPIXS XLI26'8LITPERX46%WLEPPFIFEWIHSRXLIEWFYMPXEWSTIVEXIH
| |
| ERHQEMRXEMRIHTPERX ERHVIJPIGXXLISTIVEXMRKI\TIVMIRGIEXXLITPERX
| |
| | |
| as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program in Amendment No. [###], 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.
| |
| | |
| Completion Times
| |
| : 1. 3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued)
| |
| Condition Bis entered, but continues from the time
| |
| -27)68 )<%140)
| |
|
| |
| ,..________ 1 Condition A was initially entered. If Required Action A.I is met after Condition Bis entered, Condition Bis exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not xpired.
| |
| IMMEDIATE When "Immediately" is used as a Completion Time, the COMPLETION TIME Required Action should be pursued without delay and in a controlled manner.
| |
| GRAND GULF 1.0-23 Amendment No. 120
| |
| | |
| 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.
| |
| APPLICABILITY: MODES 1 and 2.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Product of Sodium A.1 Restore (C) (E) ;::: 420 8 hours Pentaborate concentration in weight percent (C) times Boron-10 enrichment in atom percent (E) < 420 B. Sodium pentaborate B.1 Restore Volume to 8 hours solution volume 2:: 4,200 gallons.
| |
| < 4,200 gallons.
| |
| C. Sodium pentaborate C. l Restore temperature 8 hours solution temperature to 2:: 45°F and ~ 150°F.
| |
| < 45°F or > 150°F.
| |
| D. One SLC subsystem D. l Restore SLC subsystem 7 days inoperable for reasons to OPERABLE status.
| |
| other than Conditions A, B or C.
| |
| E. Two SLC subsystems E. l Restore one SLC 8 hours ~
| |
| inoperable for reasons subsystem to OPERABLE other than Conditions A, B or C.
| |
| status. I-27)68 6-'8
| |
| F. Required Action and F. l Be in MODE 3. 12 hours associated Completion Time not met.
| |
| I GRAND GULF 3.1-21 Amendment N o . ~ ~
| |
| | |
| RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.
| |
| APPLICABILITY: According to Table 3.3.1.1-1.
| |
| ACTIONS
| |
| -------------------------------------NOTE-------------------------------------
| |
| Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in 12 hours channels inoperable. trip.
| |
| OR
| |
| ---------Note--------
| |
| ~..._ _ _
| |
| -27)68 6-'8
| |
| Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.
| |
| A.2 Place associated trip 12 hours system in trip.
| |
| -27)68 6-'8
| |
| ---------NOTE--------- B.l Place channel in one 6 hours Not applicable for trip system in trip.
| |
| Functions 2.a, 2.b, 2.c, 2.d, or 2.f. OR
| |
| -27)68 6-'8
| |
| B.2 Place one trip system B. One or more Functions in trip.
| |
| with one or more required channels inoperable in both -27)68 6-'8
| |
| trip systems.
| |
| (continued)
| |
| GRAND GULF 3.3-1 Amendment N o . ~ +/--8--8
| |
| | |
| EOC - RPT I ns t rumentat i on
| |
| : 3. 3 . 4 . 1 3.3 INSTRUMENTATION 3.3. 4 .1 End of Cycle Rec ircu lati on Pump Trip (EOC - RPT) I n s t r umen ta tion LCO 3.3. 4 .1 a. Tw o channe l s p er trip sys t em f o r each EOC-RPT in s trumentation Function l i sted be low s ha l l be OPERABLE :
| |
| : 1. Turbine Sto p Va l ve (T SV ) Cl osure , Trip Oil Pressure - Low ; and
| |
| : 2. Turb ine Contro l Va l ve (T CV ) Fast Cl osure , Tr i p Oil Pressure - Low .
| |
| OR
| |
| : b. LCO 3 . 2 . 2 , "MINI MUM CRITI CAL POWER RATI O (MCPR ) , "
| |
| l imits for i n opera bl e EOC - RPT as s p ec i fi ed in t h e COLR are made app l icab l e .
| |
| AP PL I CABI L ITY: THERMAL POWER ?. 35 . 4 % RTP wi t h any recircu l at i on pump i n fast I s p eed .
| |
| AC TIONS
| |
| ---- - -- -- - --------------- - ---------- NOTE ----- - - - ------- -- - - ------ -- -------- - --
| |
| Sep ara t e Condi t ion entry is a ll owed for e a ch channe l.
| |
| CO NDITION REQ UI RED AC TIO N COMPLETI ON TI ME A. One or more re q uired A. l Res t ore channel to 72 hours ch a nn el s i no per a b l e . OPERABL E s t a t u s .
| |
| OR ~ . . . --27)68
| |
| ------ 6-'8
| |
| -----
| |
| (cont i nued )
| |
| GRAN D GULF 3 . 3 - 25 Ame ndment No . 120 , 1:ih 1-9 EOC - RPT I nstrumentation 3.3.4.1 ACTIONS CONDITION REQUI RED AC TION COMPLET I ON TI ME A. (co nt i nu ed) A. 2 ------- NOTE ------------
| |
| Not app l icable if inoperable channel is the result of an inoperable breaker .
| |
| 72 hou r s Place channel in trip .
| |
| -27)68 6-'8
| |
| ~--------,
| |
| B. One or more Functions B. l Restore EOC - RPT t rip 2 hours with EOC-RPT trip capability .
| |
| capability no t maintained .
| |
| -OR AND MCPR limit for B. 2 Apply the MCPR limit 2 hours inoperable EOC-RPT not for inoperable EOC - RPT made applicab l e . as specified in the COLR .
| |
| C. Required Ac t ion and C. l Remove the associated 4 hours associated Completion recirculation pump fast Time no t met . speed breaker from service .
| |
| -OR C. 2 Reduce THERMAL POWER to 4 hours
| |
| < 35 . 4% RTP .
| |
| GRAND GUT.F 3.3 - 26 Amendment No. 120 , ~ +/--9-+/-
| |
| | |
| ATWS-RPT Instrumentation 3.3.4.2 3.3 INSTRUMENTATION 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT} Instrumentation LCO 3.3.4.2 Two channels per trip system for each ATWS-RPT instrumentation Function listed below shall be OPERABLE:
| |
| : a. Reactor Vessel Water Level-Low Low, Level 2; and
| |
| : b. Reactor Vessel Pressure-High.
| |
| APPLICABILITY: MOOE 1.
| |
| ACTIONS
| |
| ------------------------ -------------NOTE--- ------------------------ ----------
| |
| Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME
| |
| --- 14 days A. One or more channels A.l Restore channel to inoperable. OPERABLE status.
| |
| OR 6-'8 -27)68
| |
| A.2 --------NOTE---------
| |
| Not applicable if inoperable channel is the result of an inoperable breaker.
| |
| Place channel in 14 days trip.
| |
| 6-'8 -27)68
| |
| GRANO GULF 3.3-29 Amendment No. ~
| |
| | |
| ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 -----------NOTE------------
| |
| Only applicable for Functions 3.a and 3.b.
| |
| Declare High Pressure 1 hour from discovery Core Spray (HPCS) of loss of HPCS System inoperable. initiation capability
| |
| -AND B.3 Place channel in trip. 24 hours
| |
| ~
| |
| C. As required by Required C.1 -----------NOTE------------ ~-27)68 6-'8 I Action A.1 and Only applicable for referenced in Functions 1.c, 1.d, 2.c, Table 3.3.5.1-1. and 2.d.
| |
| Declare supported 1 hour from discovery feature(s) inoperable of loss of initiation when its redundant feature capability for ECCS initiation capability feature(s) in both is inoperable. divisions
| |
| -AND (continued)
| |
| GRAND GULF 3.3-33 Amendment No. 120, 218
| |
| | |
| ECCS Instrumentation 3.3.5.1 ACTIONS CONDITION REQUIRED ACT ION COMPLETION TIME C. (continued) C.2 Restore channel to 24 hours OPERABLE status.
| |
| \
| |
| D. As required by 0.1 --------NOTE--------- ~-27)68 6-'8 I Required Action A.1 Only applicable if and referenced in HPCS pump suction is Table 3.3.5.1-1. not aligned to the suppression pool.
| |
| Declare HPCS System I hour from inoperable. discovery of loss of HPCS initiation capability AND D.2.1 Place channel in 24 hours trip.
| |
| OR 0.2.2 Align the HPCS pump
| |
| ~
| |
| 24 hours
| |
| -27)68 6-'8
| |
| suction to the suppression pool.
| |
| (continued)
| |
| GRAND GULF 3.3-34 Amendment No. 1:-2-&
| |
| | |
| ECCS Instrumentation 3.3.5.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME E. As required by Required E.1 -------------NOTE-------------
| |
| Action A.1 and Only applicable for referenced in Functions 1.e, 1.f, and 2.e.
| |
| Table 3.3.5.1-1. -----------------------------------
| |
| Declare supported 1 hour from discovery feature(s) inoperable of loss of initiation when its redundant feature capability for ECCS initiation capability feature(s) in both is inoperable. divisions
| |
| -AND E.2 Restore channel to 7 days OPERABLE status.
| |
| \ --,-27)68 6-'8 I F. As required by Required F.1 Declare Automatic 1 hour from discovery Action A.1 and Depressurization System of loss of ADS referenced in (ADS) valves inoperable. initiation capability in Table 3.3.5.1-1. both trip systems
| |
| -AND (continued)
| |
| GRAND GULF 3.3-35 Amendment No. 120, 218
| |
| | |
| ECCS Instrumentation 3.3.5.1 I-27)68 6-'8
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLE'~ION TIME F. (continued) F.2 Place channel in 96 hour ~ from trip. discovery of inoperable channel concurrent with HPCS or reactor core isolation cooling (RCIC) inoperable AND 8 days
| |
| ~
| |
| G. As required by G. l --------NOTE--------- ~......--------1 6-'8 -27)68
| |
| Required Action A.1 Only applicable for and referenced in Functions 4.c, 4.e, Tab 1e 3. 3. 5. 1-1. 4.f, 4.g, 5.c, 5.e, and 5.f.
| |
| Declare ADS valves 1 hour from inoperable. discovery of loss of ADS initiation capability in both trip systems ANO (continued)
| |
| GRAND GULF 3.3-36 Amendment No. 1-2&
| |
| | |
| ECCS Instrumentation 3.3.5.1 I6-'8 -27)68
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLE \ION TIME G. (continued) G.2 Restore channel to 96 hour ~ from OPERABLE status. discovery of inoperable channel concurrent with HPCS or RCIC inoperable AND 8 days H. Required Action and associated Completion H.1 Declare associated supported feature(s)
| |
| Inmedia~
| |
| Time of Condition B, inoperable. I6-'8 -27)68 I C, D, E, F, or G not met.
| |
| GRAND GULF 3.3-37 Amendment No. :i:-2 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.
| |
| 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 channels A.1 Enter the Condition Immediately inoperable. referenced in Table 3.3.5.3-1 for the channel.
| |
| B. As required by Required B.1 Declare RCIC System 1 hour from discovery Action A.1 and inoperable. of loss of RCIC referenced in initiation capability Table 3.3.5.3-1.
| |
| -AND B.2 Place channel in trip. 24 hours C. As required by Required C.1 Restore channel to 24 hours
| |
| '--I 6-'8 -27)68 I Action A.1 and OPERABLE status.
| |
| referenced in Table 3.3.5.3-1.
| |
| (continued)
| |
| GRAND GULF 3.3-44 Amendment No. 120, 218
| |
| | |
| RCIC System Instrumentation 3.3.5.3 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. As required by D.1 ----------NOTE----------------
| |
| Required Action A.1 Only applicable if RCIC and referenced in pump suction is not Table 3.3.5.3-1. aligned to the suppression pool.
| |
| Declare RCIC System 1 hour from discovery inoperable. of loss of RCIC initiation capability AND D.2.1 Place channel in trip. 24 hours OR 6-'8 -27)68
| |
| D.2.2 Align RCIC pump suction 24 hours 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.
| |
| GRAND GULF 3.3-45 Amendment No. 120, 218
| |
| | |
| Primary Containment and Drywell Isolation Instrumentation 3.3.6.1 3.3 INSTRUMENTATION 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation LCO 3.3.6.1 The primary containment and drywell 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
| |
| - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -NOTES - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
| |
| : 1. Penetration flow paths may be unisolated intermittently under administrative control.
| |
| : 2. Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A. l Place channel in 12 hours for channels inoperable. trip. Functions 2. b, 5.b, 5. C, and 5.d AND ~ -27)68 6-'8 I 24 hours for Functions other than Functions 2. b,
| |
| : 5. b, 5. C, and 5.d
| |
| ~
| |
| B. One or more automatic B.l Restore isolation 1 hour~
| |
| -27)68 6-'8 I Functions with capability.
| |
| isolation capability not maintained.
| |
| (continued)
| |
| GRAND GULF 3.3-48 Amendment No. 2-G-, tt:i:t"
| |
| | |
| RHR Containment Spray System Instrumentation 3.3.6.3 3.3 INSTRUMENTATION 3.3.6.3 Residual Heat Removal (RHR) Containment Spray System Instrumentation LCO 3.3.6.3 The RHR Containment Spray System instrumentation for each Function in Table 3.3.6.3-1 shall be OPERABLE.
| |
| 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 required A.1 Enter the Condition Immediately channels inoperable. referenced in Table 3.3.6.3-1 for the channel.
| |
| B. As required by B. l Declare associated 1 hour from Required Action A.I RHR containment spray discovery of and referenced in subsystem inoperable. loss of RHR Table 3.3.6.3-1. containment spray initiation capability in both trip systems ANO B.2 Place channel in 24 hours trip.
| |
| ~:,..
| |
| (continued)
| |
| Lf-27)68 6-'8 I GRANO GULF 3.3-63 Amendment No. l RHR Containment Spray System Instrumentation 3.3.6.3 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. As required by C. 1 Declare associated 1 hour from Required Action A.l RHR containment spray discovery of and referenced in subsystem inoperable. loss of RHR Table 3.3.6.3-1. containment spray initiation capability in both trip systems AND C.2 Restore channel to 24 hours OPERABLE status.
| |
| \ I
| |
| --, 6-'8 -27)68 I D. Required Action and D. l Declare associated Immediately associated Completion RHR containment spray Time of Condition B subsystem inoperable.
| |
| or C not met.
| |
| GRAND GULF 3.3-64 Amendment No. ~
| |
| | |
| Relief and LLS Instrumentation 3.3.6.5 3.3 INSTRUMENTATION 3.3.6.5 Relief and Low-Low Set (LLS} Instrumentation LCO 3.3.6.5 Two relief and LLS instrumentation trip systems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One trip system A. l Restore trip system 7 days inoperable. to OPERABLE status.
| |
| OR A.2 Declare associated relief and LLS
| |
| \
| |
| 7 days 6-'8 -27)68
| |
| valve(s} inoperable.
| |
| : 8. Required Action and B. l Be in MODE 3. 12 hours associated Completion Time of Condition A AND not met.
| |
| 8.2 Be in MOOE 4. 36 hours OR Two trip systems inoperable.
| |
| GRAND GULF 3.3-71 Amendment No. 1-2.0.
| |
| | |
| 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.
| |
| APPLICABILITY: MODES 1, 2, and 3, When the associated diesel generator (DG) is required to be OPERABLE by LCO 3.8.2, "AC Sources-Shutdown."
| |
| ACTIONS
| |
| -------------------------------------NOTE-- -----------------------------------
| |
| Separate Condition entry is allowed for each channel.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A. I Pl ace channe 1 in 24 hours inoperable. trip.
| |
| \
| |
| B. One or more Functions 8 .1 Restore actuation 1 hou~ -27)68 6-'8 I with actuation capability.
| |
| capability not maintained.
| |
| C. Required Action and C.1 Declare associated DG Irrmediately associated Completion inoperable.
| |
| Time not met.
| |
| GRAND GULF Amendment No. ~
| |
| | |
| ECCS-Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of eight safety/relief valves shall be OPERABLE.
| |
| -------------------------------------------NOTE-----------------------------------------------
| |
| Low pressure coolant injection (LPCI) subsystems may be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the residual heat removal cut in permissive pressure in MODE 3, if capable of being manually realigned and not otherwise inoperable.
| |
| APPLICABILITY: MODE 1, MODES 2 and 3, except ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.
| |
| ACTIONS
| |
| -------------------------------------------------------------------NOTE--------------------------------------------------
| |
| LCO 3.0.4.b is not applicable to HPCS.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.1 Restore low pressure 7 days injection/spray subsystem ECCS injection/spray
| |
| \
| |
| inoperable. subsystem to OPERABLE status.
| |
| (continued)
| |
| ~.--------------1
| |
| -27)68 6-'8
| |
| GRAND GULF 3.5-1 Amendment No. 169, 175, 218
| |
| | |
| ECC S Operating 3.5.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACT ION COMPLETION TIME B. High Pressure Core B.l Verify by 1 hour Spray (HPCS) System administrative means inoperable. RCIC System is OPERABLE when RCIC is required to be OPERABLE.
| |
| AND B.2 Restore HPCS System 14 days to OPERABLE status.
| |
| "' (continued)
| |
| -27)68 6-'8 I GRAND GULF 3 . 5- l a Amen dm en t No. ---+-t ECCS-Operating 3.5.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. Two ECCS injection C.1 Restore one ECCS 72 hours subsystems inoperable. injection/spray subsystem to OPERABLE
| |
| -OR One ECCS injection and one ECCS spray subsystem inoperable.
| |
| status.
| |
| \ -27)68 6-'8 I D. Required Action and D.1 -------NOTE--------
| |
| associated Completion LCO 3.0.4.a is not Time of Condition A, applicable when B, or C not met. entering MODE 3.
| |
| Be in MODE 3. 12 hours E. One ADS valve E.1 Restore ADS valve to 14 days inoperable. OPERABLE status.
| |
| ~
| |
| '--I -27)68 6-'8 I F. One ADS valve F.1 Restore ADS valve to 72 hours inoperable. OPERABLE status.
| |
| -AND -OR ~ -27)68 6-'8 I One low pressure ECCS F.2 Restore low pressure 72 hours injection/spray ECCS injection/spray subsystem inoperable. subsystem to OPERABLE status.
| |
| \ -27)68 6-'8
| |
| (continued)
| |
| I GRAND GULF 3.5-2 Amendment No. 120, 201
| |
| | |
| RCIC System 3.5.3 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), 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.
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| ACTIONS
| |
| --------------------------------------------------------------------NOTE--------------------------------------------------------
| |
| LCO 3.0.4.b is not applicable to RCIC.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. RCIC System A.1 Verify by administrative 1 hour inoperable. means High Pressure Core Spray System is OPERABLE.
| |
| -AND A.2 Restore RCIC System to OPERABLE status. 14 days B. Required Action and B.1 Be in MODE 3.
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| -'-I 12 hours
| |
| -27)68 6-'8 I associated Completion Time not met. -AND B.2 Reduce reactor steam 36 hours dome pressure to
| |
| ~ 150 psig.
| |
| GRAND GULF 3.5-10 Amendment No. 120, 175, 218
| |
| | |
| Primary Containment Air Locks 3.6.1.2 ACTIONS CONDITION REQUIRED ACT ION COMPLETION TIME C. (continued) C.3 Restore air lock to 24 hours OPERABLE status.
| |
| \
| |
| \_f6-'8 -27)68 I D. Required Action and D. l Be in MODE 3. 12 hours associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours GRAND GULF 3.6-6 Amendment No. +29-
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| | |
| PCIVs 3.6.1.3 ACTION (continued)
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| CONDITION REQUIRED ACTION COMPLETION TIME A. One or more - - - - - - - - - - - - - NOTE- - - - - - - - - - - -
| |
| penetration flow paths Relief valves are not with one PCIV required to be de-activated inoperable except due provided the relief setpoint to leakage not within is at least 23 psig and one limit. of the following criteria is met:
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| : 1. the relief valve is one-inch nominal size or less, or
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| : 2. the flow path is into a closed system whose piping pressure rating exceeds the containment design pressure rating.
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| A .1 Isolate the affected 4 hours except penetration flow path for main steam by use of at least line one closed and de-activated -27)68 6-'8
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| automatic valve, closed manual valve, 8 hours for blind flange, or main steam line check valve with flow through the valve ~ -27)68 6-'8
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| secured. ~-_ _ _ _ _ __.
| |
| JSPPS[MRK MWSPEXMSR A.2 - - - - - - - -NOTE - - - - - - - - - Once pe J Isolation devices in 31 days for high radiation areas isolation may be verified by devices outside use of administrative primary means. containment, drywel l, and steam tunnel Verify the affected penetration flow path ANO is isolated.
| |
| (continued)
| |
| GRAND GULF 3.6-10 Amendment No. -&G-, ti6=:"
| |
| | |
| PCIVs 3.6.1.3 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. One or more D. 1 Isolate the affected 24 hours penetration flow paths penetration flow path t
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| with one or more by use of at least primary containment one closed and purge valves not de-activated within purge valve automatic valve, -27)68 6-'8
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| 1eakage l i mits. closed manual valve, or b1ind fl ange.
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| AND 0.2 --------NOTE--------- Once per 31 days Isolation devices in for isolation 1' high radiation areas devices outside may be verified by primary use of administrative containment means.
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| --------------------- AND IJSPPS[MRK MWSPEXMSR Verify the affected Prior to penetration flow path entering MODE 2 is isolated. or 3 from MODE 4 if not performed within the previous 92 days for isolation devices inside primary containment AND D.3 Perform SR 3.6.1.3.5 Once per 92 days for the resilient seal purge valves closed to comply with Required Action D.1. I
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| "'l JSPPS[MRK MWSPEXMSR I (continued}
| |
| GRAND GULF 3.6-12 Amendment No. ~
| |
| | |
| LLS Valves 3.6.1.6 3.6 CONTAINMENT SYSTEMS 3.6.1.6 Low-Low Set (LLS) Valves LCO 3.6.1.6 The LLS function of six safety/relief valves shall be OPERABLE.
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| APPLICABILITY: MODES 1, 2, and 3.
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| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One LLS valve A.1 Restore LLS valve to 14 days inoperable. OPERABLE status.
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| B. Required Action and associated Completion Time of Condition A not met.
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| B.1 -------NOTE--------
| |
| LCO 3.0.4.a is not applicable when entering MODE 3.
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| ' i I-27)68 6-'8 I Be in MODE 3. 12 hours C. Two or more LLS valves C.1 Be in MODE 3. 12 hours inoperable.
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| -AND C.2 Be in Mode 4. 36 hours GRAND GULF 3.6-20 Amendment No. 120, 201
| |
| | |
| RHR Containment Spray System 3.6.1.7 3.6 CONTAINMENT SYSTEMS 3.6.1.7 Residual Heat Removal (RHR) Containment Spray System LCO 3.6.1.7 Two RHR containment spray subsystems shall be OPERABLE.
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| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR containment A.1 Restore RHR 7 days spray subsystem containment spray inoperable. subsystem to OPERABLE ),
| |
| status.
| |
| I I
| |
| -27)68 6-'8 I B. Two RHR containment B.1 Restore one RHR 8 hours spray subsystems containment spray inoperable. subsystem to OPERABLE status.
| |
| C. Required Action and C.1 -------NOTE-------
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| associated Completion LCO 3.0.4.a is not Time not met. applicable when entering MODE 3.
| |
| Be in MODE 3. 12 hours GRAND GULF 3.6-22 Amendment No. 120, 201
| |
| | |
| 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 OPERABLE status.
| |
| I B. Required Action and B.1 ------NOTE-------- I-27)68 6-'8 I associated Completion LCO 3.0.4.a is not Time of Condition A applicable when not met. entering MODE 3.
| |
| 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 GRAND GULF 3.6-31 Amendment No. 120, 201
| |
| | |
| SPMU System 3.6.2.4 3.6 CONTAINMENT SYSTEMS 3.6.2.4 Suppression Pool Makeup (SPMU) System LCO 3.6.2.4 Two SPMU subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACT ION COMPLETION TIME A. Upper containment pool A.1 Restore upper 4 hours water level not within containment pool 1imi t. water level to within 1imi t.
| |
| B. Upper containment pool B.1 Restore upper 24 hours water temperature not containment pool within limit. water temperature to within 1imit.
| |
| C. One SPMU subsystem C.l Restore SPMU 7 days inoperable for reasons subsystem to OPERABLE other than Condition A or B.
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| status.
| |
| t -27)68 6-'8
| |
| D. Required Action and D. I Be in MODE 3. 12 hours associated Completion Time not met. AND D.2 Be in MODE 4. 36 hours GRAND GULF 3.6-33 Amendment No. 120.
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| | |
| Drywell Air Lock 3.6.5.2 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME C. Drywell air lock C. l Verify a door is 1 hour inoperable for reasons closed.
| |
| other than Condition A or B. AHll C.2 Restore air lock to 24 hours OPERABLE status. J\
| |
| y -27)68 6-'8 I D. Required Action and 0.1 Be in MODE 3. 12 hours associated Completion Ti111e not met. M!Q 0.2 Be in MODE 4. 36 hours GRAND GULF 3.6-56 Amendment No. m, 4 I
| |
| Orywell Isolation Valves 3.6.5.3 3.6 CONTAINMENT SYSTEMS 3.6.5.3 Drywell Isolation Valves LCO 3.6.5.3 Each drywell isolation valve, except for Drywell Vacuu Relief Systea valves, shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS
| |
| -------------------------------------NOTES ------------------------------------
| |
| : 1. Penetration flow paths uy 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 drywell isolation valves.
| |
| CONDITION REQUIRED ACTION COMPLETION TIME A. One or 1K>re penetration flow paths with one drywell isolation valve inoperable.
| |
| A.I Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual 8
| |
| t hours
| |
| .....-27)68
| |
| _ _ _6-'8
| |
| __
| |
| valve, blind flange, or check valve with flow through the valve secured.
| |
| (continued)
| |
| GRAND GULF 3.6-58 Amendment No. 29, ~
| |
| | |
| SSW System and UHS 3.7.1 3.7 PLANT SYSTEMS 3.7.1 Standby Service Water (SSW) System and Ultimate Heat Sink (UHS)
| |
| LCO 3.7.1 Division 1 and 2 SSW subsystems and the UHS shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One UHS cooling tower A. l Restore UHS cooling 7 days with one cooling tower tower fan to OPERABLE
| |
| ~-27)68 6-'8
| |
| fan inoperable. status.
| |
| B. One UHS cooling tower B. l Declare associated Immediately with two cooling tower SSW subsystem fans inoperable. inoperable.
| |
| C. UHS basin level not C.1 Restore UHS basin 72 hours within limit. level to within limit.
| |
| (continued)
| |
| GRAND GULF 3.7-1 Amendment No. 1-20.
| |
| | |
| SSW System and UHS 3.7.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. One SSW subsystem --------------NOTES----------------------
| |
| inoperable. 1. Enter applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources -
| |
| Operating," for diesel generator made inoperable by SSW.
| |
| : 2. Enter applicable Conditions and Required Actions of LCO 3.4.9, "Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown," for RHR shutdown cooling subsystem made inoperable by SSW.
| |
| D.1 Restore SSW subsystem to 72 hours OPERABLE status.
| |
| ~'
| |
| E. Required Action and E.1 ------------NOTE-------------- Lf-27)68 6-'8 I associated Completion LCO 3.0.4.a is not Time of Condition A, C, applicable when entering or D not met. MODE 3.
| |
| Be in MODE 3. 12 hours (continued)
| |
| GRAND GULF 3.7-2 Amendment No. 120, 219
| |
| | |
| AC SourcesOperating 3.8.8
| |
|
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 Restore required offsite 72 hours circuit to OPERABLE
| |
| -AND~ I status. -27)68 6-'8
| |
| 24 hours from discovery of two divisions with no offsite power I
| |
| \
| |
| B. One required DG B.1 Perform SR 3.8.1.1 for 1 hour
| |
| \...r-27)68 6-'8 I inoperable for reasons OPERABLE required other than Condition F. offsite circuit(s). -AND Once per 8 hours thereafter
| |
| -AND 4 hours from B.2 Declare required discovery of feature(s), supported by Condition B the inoperable DG, concurrent with inoperable when the inoperability of redundant required redundant required feature(s) are inoperable. feature(s)
| |
| -AND (continued)
| |
| GRAND GULF 3.8-2 Amendment No. 120, 151, 226
| |
| | |
| AC SourcesOperating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) 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 for 24 hours OPERABLE DG(s).
| |
| -AND B.4 Restore required DG to 72 hours from OPERABLE status. discovery of an inoperable Division 3 DG
| |
| -AND~ -27)68 6-'8 I 14 days
| |
| ~
| |
| I
| |
| "--I-27)68 6-'8 I 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) are concurrent with inoperable. inoperability of redundant required feature(s)
| |
| -AND 24 hours C.2 Restore one required
| |
| \~
| |
| offsite circuit to OPERABLE status.
| |
| (continued)
| |
| I-27)68 6-'8 I GRAND GULF 3.8-3 Amendment No. 120, 151, 226
| |
| | |
| AC Sources ² Operating 3.8.1 ACTIONS (continued)
| |
| CONDITION REQUIRED ACTION COMPLETION TIME D. One required offsite -----------NOTE--------------
| |
| circuit inoperable for Enter applicable Conditions reasons other than and Required Actions of Condition F. LCO 3.8.7, "Distribution Systems C Operating," when any
| |
| -AND required division is de-energized as a result of One required DG Condition D.
| |
| inoperable for reasons -----------------------------
| |
| other than Condition F. D.1 Restore required 12 hours offsite circuit to
| |
| ~
| |
| OPERABLE status.
| |
| -OR
| |
| -27)68 6-'8 I D.2 Restore required DG 12 hours to OPERABLE status.
| |
| E. Two required DGs inoperable.
| |
| E.1 Restore one required DG to OPERABLE status.
| |
| ' --, -27)68 6-'8 I 2 hours
| |
| -OR 24 hours if Division 3 DG is inoperable F. One automatic load F.1 Restore automatic 24 hours sequencer inoperable. load sequencer to ,r,..
| |
| OPERABLE status.
| |
| G. Required Action and G.1 --------NOTE---------
| |
| LI -27)68 6-'8 I associated Completion LCO 3.0.4.a is not Time of Condition A, applicable when B, C, D, E, or F not entering MODE 3.
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| met. ---------------------
| |
| Be in MODE 3. 12 hours (continued)
| |
| GRAND GULF 3.8-4 Amendment No. 120, 201
| |
| | |
| DC Sources-Operating 3.8.4
| |
| --- 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources-Operating LCO 3.8.4 The Division 1, Division 2, and Division 3 DC electrical power subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required battery ------------NOTE-------------
| |
| charger inoperable. Entry into MOOE 1, 2 or 3 is not allowed, except during power reductions.
| |
| A.1 Verify battery cell 1 hour parameters meet
| |
| -* Table 3.8.6-1 Category A limits.
| |
| ANO Once per 8 hours thereafter B. Required Action and B. l Declare associated Invnediately associated Completion battery inoperable.
| |
| Time of Condition A not met.
| |
| C. Division 1 or 2 DC C.l Restore Division I 2 hours and 2 DC electrical t
| |
| electrical power subsystem inoperable power subsystems to for reasons other than OPERABLE status.
| |
| Condition A. -27)68 6-'8
| |
| (continued)
| |
| GRAND GULF 3.8-26 Amendment No. l-2G-
| |
| | |
| Distribution SystemsOperating 3.8.7 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Distribution SystemsOperating LCO 3.8.7 Division 1, Division 2, and Division 3 AC and DC electrical power distribution subsystems shall be OPERABLE.
| |
| APPLICABILITY: MODES 1, 2, and 3.
| |
| NOTE Division 3 electrical power distribution subsystems are not required to be OPERABLE when High Pressure Core Spray System is inoperable.
| |
| ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Division 1 A.1 Restore Division 1 and 2 8 hours or 2 AC electrical power AC electrical power
| |
| ~
| |
| distribution distribution subsystems to subsystem(s) OPERABLE status.
| |
| inoperable. -27)68 6-'8
| |
| B. One or more Division 1 B.1 Restore Division 1 and 2 2 hours or 2 DC electrical power DC electrical power distribution subsystem(s) inoperable.
| |
| distribution subsystems to OPERABLE status.
| |
| t -27)68 6-'8
| |
| (continued)
| |
| I GRAND GULF 3.8-38 Amendment No. 120, 226
| |
| | |
| Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Control Room Envelope Habitability Program (continued)
| |
| : 1. Plant maintenance activities such as modifications, rework, and preventive maintenance tasks on components that could affect the CRE shall be controlled under fleet, plant and system specific procedures to ensure that the CRE boundary is not degraded by such activities.
| |
| : 2. Testing of CRFA system sealing areas shall be performed following maintenance activities (rework and preventative) and periodically to ensure that the areas of negative pressures do not leak bypassing emergency filtration system components.
| |
| : 3. Fire damper inspection procedures that require opening of duct panels and doors shall ensure that upon restoration no leakage path exists.
| |
| : 4. The remainder of ducting components such as plenum access doors, duct access doors (rectangular and round), flex connections (ventglass, etc), plugs, and patches will be maintained per paragraph b.
| |
| : 5. An assessment of the CRE Boundary will be conducted at a frequency in accordance with the Surveillance Frequency Control Program. The results of assessing items 1 through 4 shall be trended and used as part of the assessment of the CRE boundary as indicated in paragraph c.
| |
| : e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of OBA consequences.
| |
| Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
| |
| : f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and assessing the CRE boundary as required by paragraphs c and d, respectively.
| |
| -27)68 6-'8 463+6%1 GRAND GULF 5.0-16b
| |
| ==-- ,
| |
| Amendment No. 178, 219, 227
| |
| | |
| Attachment 3 GNRO-2023/00014 Technical Specifications Bases Page Markups For Information Only (45 pages follow)
| |
| | |
| INSERT 3 or in accordance with the Risk Informed Completion Time Program INSERT 4 Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
| |
| INSERT 5 Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. This Completion Time is modified by a Note to clarify that the Risk Informed Completion Time Program is not Applicable to a Required Action associated with a Condition that could represent a loss of safety function as defined in TS 5.5.10.
| |
| | |
| SLC System B 3.1.7 BASES ACTIONS B.1 If the volume of the sodium pentaborate solution is less than 4,200 gallons, the volume must be restored to greater than or equal to 4,200 gallons within 8 hours. When in Condition B.1, it is not necessary to enter Condition E for both SLC subsystems inoperable. The subsystems are capable of performing their original design basis function. Because of the low probability of an ATWS event and that the SLC System capability still exists for vessel injection under this condition, the allowed Completion Time of 8 hours is acceptable and provides adequate time to restore the volume to within limits.
| |
| C.1 If the temperature of the sodium pentaborate solution is less than 450 F or greater than 150 0 F, the temperature must be restored to within limits within 8 hours. When in Condition C.1, it is not necessary to enter Condition E for both SLC subsystems inoperable. The subsystems are capable of performing their original design basis function. Because of the low probability of an ATWS event and that the SLC System capability still exists for vessel injection under this condition, the allowed Completion Time of 8 hours is acceptable and provides adequate time to restore the temperature to within limits.
| |
| -27)68
| |
| D.1 If one SLC subsystem is inoperable for reasons other than Conditions A, B or C, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE subsystem is adequate to perform the shutdown function. However, the overall reliability 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 performing the intended SLC System function and the low probability of a Design Basis Accident (DBA) or severe transient occurring concurrent with the failure of the Control Rod Drive System to shut down the plant.
| |
| E.1 If both SLC subsystems are inoperable for reasons other than Conditions A, B or C, 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 continued)
| |
| GRAND GULF B 3.1-39 LBDCR 16270
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES ACTIONS Times, specifies that once a Condition has been entered, (continued) 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 Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for 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.
| |
| A.1 and A.2 Because of the diversity of sensors available to provide ______
| |
| trip signals and the redundancy of the RPS design, an -27)68
| |
| allowable out of service time of 12 hours has been be acceptable (Ref. 9 and 15) to permit restoratio inoperable channel to OPERABLE status. However, is out of service time is only acceptable provided the as ociated Function's inoperable channel is in one trip stem and the Function still maintains RPS trip capabilit (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 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 (or trip system) in trip (e.g.,
| |
| as in the case where placing the inoperable channel in trip would result in a full scram), Condition D must be entered and its Required Action taken.
| |
| As noted, Action A.2 is not applicable to APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f. Inoperability of one required APRM/OPRM channel affects both trip systems. For that condition, Required Action A.1 must be satisfied, and is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure.
| |
| (continued)
| |
| GRANO GULF B 3.3-19 LBDCR 18827
| |
| | |
| RPS Instrumentation B 3.3.1.1 BASES ACTIONS B.1 and B.2 (continued) of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.
| |
| < -27)68
| |
| Alternately, if it is not desired to place the inoperable channels (or one trip system) in trip (e.g., as in the case where placing the inoperable channel or associated trip system in trip would result in a scram or RPT), Condition D must be entered and its Required Action taken.
| |
| As noted, Condition B is not applicable to APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f. Inoperability of one required APRM/OPRM channel affects both trip systems and is not associated with a specific trip system, as are the APRM 2-Out-Of-4 Voter and other non-APRM/OPRM channels for which Condition B applies. For an inoperable APRM/OPRM channel, Required Action A.1 must be satisfied, and is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of more than one required APRM/OPRM channel of the same trip function results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel. Because Conditions A and C provide Required Actions that are appropriate for the inoperability of APRM Functions 2.a, 2.b, 2.c, 2.d, and 2.f, and these functions are not associated with specific trip systems as are the APRM 2-Out-Of-4 Voter and other non-APRM channels, Conditions B does not apply.
| |
| C.1 Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped 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 (continued)
| |
| GRAND GULF B 3.3-21 LBDCR 16277
| |
| | |
| EOC-RPT Instrumentation B 3.3.4.1 BASES ACTIONS channels. As such, a Note has been provided that allows (continued) separate Condition entry for each inoperable EOC-RPT instrumentation channel.
| |
| A.1 and A.2 With one or more channels inoperable, but with EOC-RPT trip capability maintained (refer to Required Action B.1 Bases),
| |
| the EOC-RPT System is capable of performing the intended function. However, the reliability and redundancy of the EOC-RPT instrumentation is reduced such that a single failure in the remaining trip system could result in the inability of the EOC-RPT System to perform the intended function. Therefore, only a limited time is allowed to restore compliance with the LCO. 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 an EOC-RPT, 72 hours is allowed to restore the inoperable channels (Required Action A.1) or apply the EOC-RPT inoperable MCPR Limit.
| |
| Alternately, the inoperable channels may be placed in trip (Required Action A.2) since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to
| |
| -27)68 continue. As noted in Required Action A.2, placing the c el in trip with no further restrictions is not allowed if the
| |
| * erable channel is the result of an inoperable breaker, si this may not adequately compensate for the inoperable breaK (e.g., the breaker may be inoperable such that it will not open 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 EOC-RPT), or if the inoperable channel is the result of an inoperable breaker, Condition C must be entered and its Required Actions taken.
| |
| B.1 and B.2 Required Actions B.1 and B.2 are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in the Function not maintaining EOC - RPT trip capability. A Function is considered to be maintaining EOC-RPT trip capability when continued GRAND GULF B 3.3-72 LDC 01128
| |
| | |
| ATWS-RPT Instrumentation B 3.3.4 . 2 BASES ACTIONS Actions of the Condition continue to apply for each (continued) additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable ATWS-RPT 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 ATWS-RPT instrumentation channel.
| |
| A.land A.2 With one or more channels inoperable, but with ATWS-RPT capability for each Function maintained (refer to Required Action 8.1 and C.l Bases), the ATWS-RPT System is capable of performing the intended function. However, the 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.I) . 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 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 -27)68
| |
| inoperable breaker, since this may not adequat compensate for the inoperable breaker (e.g., the brea may be inoperable such that is will not open). fit is not desirable to place the channel in trip (e.g., as in the case where placing the inoperable channel would result in an ATWS-RPT), or if the inoperable channel is the result of an inoperable breaker, Condition D must be entered and its Required Actions taken.
| |
| (continued)
| |
| GRANO GULF B 3.3-82 Revision No. +
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS 8.1, 8.2, and 8.3 (continued} -27)68
| |
| Because of the diversity of sensors avail ble to provide initiation signals and the redundancy of he ECCS design, an allowable out of service time of 24 hou r. has been shown to be acceptable (Ref. 4} to permit resto tion 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 8.3. 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 H must be entered and its Required Action taken.
| |
| C.l and C.2 Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the same Function (or in some cases, within the same variable} result in redundant automatic initiation capability being lost for the feature(s}. Required Action C.l features would be those that are initiated by Functions 1.c, l.d, 2.c, and 2.d (i.e., low pressure ECCS}.
| |
| For Functions l*.c and 2.c, redundant automatic initiation capability is lost if the Function 1.c and Function 2.c channels are inoperable. For Functions l.d and 2.d, redundant automatic initiation capability is lost if two Function 1.d channels in the same trip system and two Function 2.d channels in the same trip system (but not necessarily the same trip system as the Function 1.d channels} are inoperable. Since each inoperable channel would have Required Action C.1 applied separately (refer to ACTIONS Note}, each inoperable channel would only require the affected portion of the associated Division to be declared inoperable. However, since channels in both Divisions are inoperable, and the Completion Times started concurrently for the channels in both Divisions, this results in the affected portions in both Divisions being concurrently declared inoperable. For Functions 1.c and 2.c, the affected portions of the Division are LPCI A (continued)
| |
| GRAND GULF B 3.3-112 Revision No.
| |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS C.l and C.2 (continued)
| |
| -27)68
| |
| Because of the diversity to provide initiation signals and the redundancy oft ECCS design, an allowable out of service time of 24 hours as been shown to be acceptable (Ref. 4) to permit restor ion 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 would not necessarily result in a safe state for the channel in all events.
| |
| D.l, D.2.1, and 0.2.2 Required Action D.l 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 component initiation capability for the HPCS System. Automatic component initiation capability is lost if two Function 3.d channels or two Function 3.e channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour allowance of Required Actions 0.2.1 and 0.2.2 is not appropriate and the HPCS System must be declared inoperable within 1 hour after I '
| |
| discovery of loss of HPCS initiation capability. As noted, the Required Action is only applicable if the HPCS 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 0.1, the Completion Time only begins upon discovery that the HPCS 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.
| |
| (continued)
| |
| GRAND GULF B 3.3-114 Revision No.
| |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS D.l, D.2.1, and D.2.2 (continued) -27)68
| |
| Because of the diversity of sensors availa e to provide initiation signals and the redundancy of e ECCS design, an allowable out of service time of 24 hour has been shown to be acceptable (Ref. 4) to permit restor. ion of any [MXLMR LSYVW inoperable channel to OPERABLE status. If the inopera e channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed n the tripped condition per Required Action 0.2.1 or the suction source must be aligned to the suppression pool er Required Action D.2.2. Placing the inoperable channel in trip performs the intended function of the channel (shifting the suction source to the suppression pool). Performance of either of these two Required Actions will allow operation to continue. If Required Action 0.2.1 or Required Action 0.2.2 is performed, measures should be taken to ensure that the HPCS System piping remains filled with water. Alternately, if it is not desired to perform Required Actions 0.2.1 and 0.2.2, Condition H must be entered and its Required Action taken.
| |
| E. l and E. 2 Required Action E.l is intended \o ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the LPCS and LPCI Pump Discharge Flow-Low (Bypass) Functions result in redundant automatic initiation capability being lost for the feature(s). For Required Action E.l, the features would be those that are initiated by Functions l.e, l.f, and 2.e (e.g., low pressure ECCS).
| |
| Redundant automatic initiation capability is lost if three of the four channels associated with Functions 1.e, l.f, and 2.e are inoperable. Since each inoperable channel would have Required Action E.l 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.
| |
| (continued)
| |
| GRAND GULF B 3.3-115 Revision No.
| |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS E.I and E.2 (continued) protection and required flow. Furthermore, other ECCS pumps would be sufficient to complete the assumed safety function if no additional single failure were to occur. The 7 day Completion Time of Required Action E.2 to restore the inoperable channel to OPERABLE status is reasonable based on the remaining capability of the associated ECCS subsystems, the redundancy available in the ECCS design, and the low probability of a OBA occurring during the allowed out of service time. 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 not necessarily result in a safe state for the channel in all events. ~
| |
| f.l and f.Z ~-27)68
| |
| Required Action F.l is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within similar ADS trip system Functions result in automatic initiation capability being lost for the ADS.
| |
| Automatic initiation capability is lost if either (a) one or more Function 4.a channel(s) and one or more Function 5.a channel(s) are inoperable and untripped, (b) one Function 4.b channel(s) and one Function 5;b channel are inoperable and untripped, or (c) one Function 4.d channel and one Function 5.d channel 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 in both trip systems.
| |
| 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.l, 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 (continued)
| |
| GRANO GULF B 3.3-117 Revision No. -e--
| |
| | |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS F.l and F.2 (continued) paragraph above . The I 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.
| |
| -27)68 Because of the diversity of sensors available to provide i itiation signals and the redundancy of the ECCS design, an al able out of service time of 8 days has been shown to be accep able (Ref. 4) to permit restoration of any inoperable channe to OPERABLE status if both HPCS and RCIC are OPERABLE. If either HPCS or RCIC is inoperable, the time is shortened to 96 hours . he status of HPCS or RCIC changes such that the Completio---~-ft changes from 8 days to 96 hours, the 96 hours begins upon discov HPCS or RCIC in operability. However, total time for an inopera ~-, _ _..------
| |
| -27)68
| |
| untripped channel cannot exceed 8 days. If the status of .....__ _ ___,
| |
| HPCS or RCIC changes such that the Completion Time changes from 96 hours to 8 days, the "time zero" for beginning the 8 day "clock" begins upon discovery of the inoperable, untripped channel. 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 F.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 H must be entered and its Required Action taken.
| |
| G.l and G.2 Required Action G.l is intended to ensure that appropriate actions are taken if multiple, inoperable channels within similar ADS trip system Functions result in automatic initiation capability being lost for the ADS. Automatic initiation capability is lost if either (a) one Function 4.c channel and one Function 5.c channel are inoperable, (b) one or more Function 4.e channels and one or more Function 5.e channels are inoperable, (c) one or more Function 4.f channels and one or more Function 5.e channels are (continued)
| |
| GRAND GULF B 3.3-118 Revision No.
| |
| ECCS Instrumentation B 3.3.5.1 BASES ACTIONS G.l and G.2 {continued) inoperable, or {d) one or more Function 4.g channels and one or more Function 5.f channels are inoperable.
| |
| In this situation {loss of automatic initiation capability),
| |
| the 96 hour or 8 day allowance, as applicable, of Required Action G.2 is not appropriate, and all ADS valves must be declared inoperable within 1 hour after discovery of loss of ADS initiation capability in both trip systems. The Note to Required Action G.1 states that Required Action G.l is only applicable for Functions 4.c, 4.e, 4.f, 4.g, 5.c, 5.e, and 5.f. Required Action G.1 is not applicable to Functions 4.h and 5.g {which also require entry into this Condition if a channel in these Functions is inoperable),
| |
| since they are the Manual Initiation Functions and are not assumed in any accident or transient analysis. Thus, a total loss of manual initiation capability for 96 hours or 8 days {as allowed by Required Action G.2) is allowed.
| |
| 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 G.l, the Completion Time only begins upon discovery that the ADS cannot be automatically I initiated due to inoperable channels within similar ADS trip system Functions, as described in the paragraph above. The I 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.
| |
| -27)68
| |
| Becau f the diversity of sensors available to provide initiation
| |
| * nals and the redundancy of the ECCS design, an allowable out o rvice time of 8 days has been shown to be acceptable {Ref. 4) ermit restoration of any inoperable channel to OPERABLE statu both HPCS and RCIC are OPERABLE {Required Action G.2 . If either HPCS or RCIC is inoperable, the time is reduced to 96 hours. If the status of HPCS or RCIC changes such that the Comple on Time changes from 8 days to 96 hours, the 96 hours egins upon discovery of HPCS or RCIC inoperability. Howe er, total time for an inoperable channel cannot exceed 8 ays. If the status of HPCS or RCIC changes such that the Co pletion Time changes from 96 hours to 8 days, the "time zero* for
| |
| -27)68
| |
| (continued}
| |
| GRAND GULF B 3.3-119 Revision No.
| |
| RCIC System Instrumentation B 3.3.5.3 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 Water Level - Low Low, Level 2 channels 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.
| |
| -27)68
| |
| 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 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)
| |
| GRAND GULF B 3.3-130 LBDCR 18128
| |
| | |
| RCIC System Instrumentation B 3.3.5.3 BASES ACTIONS D.1, D.2.1, and D.2.2 (continued)
| |
| 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. -27)68
| |
| 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 the intended function of the channel (shifting the suction source to the suppression pool). [MXLMR LSYVW Alternatively, Required Action D.2.2 allows the manual alignment of the RCIC suction to the suppression pool, which also performs the intended function. If Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the RCIC System piping remains filled with water. If it is not desired to perform Required Actions D.2.1 and D.2.2, Condition E must be entered and its Required Action taken.
| |
| E.1 With any Required Action and associated Completion Time not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared inoperable immediately.
| |
| SURVEILLANCE As noted in the beginning of the SRs, the SRs for each RCIC REQUIREMENTS System instrumentation Function are found in the SRs column of Table 3.3.5.3-1.
| |
| The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated (continued)
| |
| GRAND GULF B 3.3-132 LBDCR 18128
| |
| | |
| Primary Containment and Drywell Isolation Instrumentation B 3.3.6.1 BASES (continued)
| |
| ACTIONS The ACTIONS are modified by two Notes. Note 1 allows penetrations flow path(s) to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment or drywell isolation is indicated.
| |
| Note 2 has been provided to modify the ACTIONS related to isolation 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 Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable primary containment isolation 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 primary containment isolation instrumentation channel.
| |
| A .1 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, has been shown to be acceptable (Refs. 5 and 6) to permit restoration of any inoperable channel to OPERABLE status. Functions that share common instrumentation with the RPS have a 12 hour allowed out of service time consistent with the time provided for the associated RPS instrumentation channels. This out of.. ..-------
| |
| service time is only acceptable provided the -27)68
| |
| Function is still maintaining isola
| |
| * a ility (refer to Required Action B.l Bases). he 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.l. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a continued GRAND GULF B 3.3-162 LDC 03033
| |
| | |
| RHR Containment Spray System Instrumentation B 3.3.6.3 ACTIONS A.l (continued)
| |
| Condition specified in the table is Function dependent.
| |
| Each time a required channel is discovered inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.
| |
| B.l and B.2 Required Action B.l is intended to ensure appropriate actions are taken if multiple, inoperable, untripped channels within the same Fun~tion result in a complete loss of automatic initiation capability for the RHR Containment Spray System. Automatic initiation capability is lost if one Function 1 channel in both trip systems is inoperable and untripped, or one Function 3 channel in both trip systems is inoperable and untripped. In this situation (loss of automatic initiation capability), the 24 hour allowance of Required Action B.2 is not appropriate and both RHR Containment Spray subsystems, made inoperable by RHR Containment Spray System instrumentation, must be declared inoperable within 1 hour after discovery of loss of RHR Containment Spray System initiation capability for both trip systems.
| |
| The Completion Time is intended to a11ow 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 8.1, the Completion Time only begins upon discovery that the RHR Containment Spray System cannot be automatically initiated due to inoperable, untripped channels within the same Function, as described in the paragraph above. The 1 hour Completion Time from discovery
| |
| ~f l~ss. of initiation capability* is acceptable because it -
| |
| minimizes risk while allowing time for restoration or tripping of channels. - - - -- - - -
| |
| -27)68 Because of the redundancy of sensors available to provide initiation signals, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 3) to permi 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
| |
| {continued)
| |
| GRAND GULF B 3.3-190 Revision No. -e-
| |
| | |
| RHR Containment Spray System Instrumentation B 3.3.6.3 BASES ACTIONS B.l and B.2 (continued) 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 the 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 D must be entered and its Required Action taken.
| |
| C.l and C.2 Required Action C.l is intended to ensure appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in automatic initiation capability being lost for the RHR Containment Spray System. Automatic initiation capability is lost if two Function 2 channels or two Function 4 channels are inoperable. In this situation (loss of automatic initiation capability), the 24 hour allowance of Required Action C.2 is not appropriate and both of the associated RHR Containment Spray subsystems must be declared inoperable within 1 hour after discovery of loss of RHR Containment Spray System 11 initiation capability for both trip systems.
| |
| 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.l, the Completion Time only begins upon discovery that the RHR Containment Spray System cannot be automatically initiated due to two inoperable channels within 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 of channels. -27)68
| |
| Because of the redundancy of sensors available to provide initiation signals, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 3) to perm*
| |
| restoration of any inoperable channel to OPERABLE status.
| |
| If the inoperable channel cannot be restored to OPERABLE (continued}
| |
| GRAND GULF B 3.3-191 Rev i s ion No.
| |
| Relief and LLS Instrumentation B 3.3.6.5 BASES APPLICABILITY overpressure limit cannot be approached by assumed (continued} operational transients or accidents. Thus, pressure relief, associated relief, and LLS instrumentation are not required.
| |
| ACTIONS A.I and A.2
| |
| -27)68
| |
| Because the failure of any reac ors eam ome pressure instrument channels in one tri system will not prevent the associated S/RV from performin its relief and LLS function, 7 days is allowed to restore trip system to OPERABLE status (Required Action A.l). In this condition, the remaining OPERABLE trip system is adequate to perform the relief and LLS initiation function. However, the overall reliability is reduced because a single failure in the OPERABLE trip system could result in a loss of relief or LLS function.
| |
| Alternatively, declaring the associated relief and LLS valve(s) inoperable (Required Action A.2) is also acceptable since the Required Actions of the respective LCOs (LCO 3.4.4 and LCO 3.6.1.6) provide appropriate actions for the inoperable components.
| |
| The 7 day Completion Time is considered appropriate for the relief and LLS function because of the redundancy of sensors available to provide initiation signals and the redundancy of the relief and LLS design. In addition, the probability of multiple relief or LLS instrumentation channel failures, which renders the remaining trip system inoperable, occurring together with an event requiring the relief or LLS function during the 7 day Completion Time is very low.
| |
| B.l and B.2 If the inoperable trip system is not restored to OPERABLE status within 7 days, per Condition A, or if two trip systems are inoperable, then 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 MOOE 3 within 12 hours and to MOOE 4 within 36 hours.
| |
| The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions
| |
| {continued)
| |
| GRAND GULF B 3.3-211 Revision No . .Q-
| |
| | |
| LOP Instrumentation B 3.3.8 . 1 BASES ACTIONS such. a Note has been provided that allows separate (continued) Condition entry for each inoperable LOP instrumentation channel.
| |
| A. l
| |
| -27)68
| |
| Because of the redundancy of sensors ava* able to provide initiation signals, 24 hours is allowed o 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.l. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to acco11111odate 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 channel in trip would result in a DG initiation), Condition B must be entered and its Required Action taken.
| |
| The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 24 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
| |
| B.1 Required Action 8.1 is intended to ensure that appropriate actions are taken if inoperable untripped channels result in the Function not maintaining LOP actuation capability for the associated Oeisel Generator (DG). A Function is considered to not be maintaining actuation capability when sufficient channels are not OPERABLE or in trip (or the associated trip system is in trip), such that the given Function will not actuate the associated DG on a valid signal. For Functions I.a, l.c, 2.a, 2.c, 2.d, and 2.e with one-out-of-two taken twice logic, this would require one of the trip systems to have at least two channels inoperable and not in trip. For Functions l.b, l.d, and 2.b this would require one trip system to have the required channel inoperable and not in trip.
| |
| (continued)
| |
| GRAND GULF B 3.3-229 Revision No . ..Q-
| |
| | |
| ECCS ² Operating B 3.5.1 BASES (continued)
| |
| ACTIONS A draining event is a slow evolution when compared to a design basis LOCA assumed to occur at full power, and thus there is adequate time to take manual actions (hours versus minutes). TS 3.5.2, Action E, prohibits plant conditions that could result in Drain Times less than one hour.
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| Therefore, there is sufficient time for the licensed operators to take manual action to stop the draining event, and to manually start an ECCS injection/spray subsystem or the additional method of water injection.
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| Consequently, there is no need for automatic initiation of ECCS to respond to an unexpected draining event in Mode 4 or 5. Automatic initiation of an ECCS injection/spray subsystem, with injection rates of thousands of gpm, may be undesirable as it can lead to overflowing the RPV cavity.
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| A Note prohibits the application of LCO 3.0.4.b to an inoperable HPCS subsystem. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable HPCS subsystem 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.
| |
| -27)68
| |
| A.1 If any one low pressure ECCS injection/spray subsystem is inoperable, the inoperable 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. 12) that evaluated the impact on ECCS availability by assuming that 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).
| |
| (continued)
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| GRAND GULF B 3.5-6 LBDCR 18128
| |
| | |
| ECCS Operating B 3.5.1 BASES ACTIONS B.1 and B.2 -27)68
| |
| If the HPCS System is inoperable, and the RCIC System is verified to be OPERABLE (when RCIC is required to be OPERABLE), the HPCS System must be restored to OPERABLE status within 14 days. In this Condition, adequate core cooling is ensured by the OPERABILITY of the redundant and diverse low pressure ECCS injection/spray subsystems in conjunction with the ADS. Also, the RCIC System will automatically provide makeup water at most reactor operating pressures. Verification of RCIC OPERABILITY within 1 hour is therefore required when HPCS is inoperable and RCIC is required to be OPERABLE. This may be performed by an administrative check, by examining logs or other information, to determine if RCIC is out of service for maintenance or other reasons. It is not necessary to perform the Surveillances needed to demonstrate the OPERABILITY of the RCIC System. However, if the OPERABILITY of the RCIC System cannot be verified and RCIC is required to be OPERABLE, Condition D must be immediately entered. If a single active component fails concurrent with a design basis LOCA, there is a potential, depending on the specific failure, that the minimum required ECCS equipment will not be available. A 14 day Completion Time is based on the results of a reliability study (Ref. 12) and has been found to be acceptable through operating experience.
| |
| (continued)
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| GRAND GULF B 3.5-6a LBDCR 18128
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| | |
| (&&6 ² 2SHUDWLQJ
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| B 3.5.1 BASES ACTIONS C.1 (continued) -27)68
| |
| With two ECCS injection subsystems inoperable or one ECCS injection and one ECCS spray subsystem inoperable, at least one ECCS injection/spray subsystem must be restored to OPERABLE status within 72 hours. In this Condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced in this Condition because a single failure in one of the remaining OPERABLE subsystems concurrent with a design basis LOCA may result in the ECCS not being able to perform its intended safety function. Since the ECCS availability is reduced relative to Condition A, a more restrictive Completion Time is imposed. The 72 hour Completion Time is based on a reliability study, as provided in Reference 12.
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| D.1 If any Required Action and associated Completion Time of Condition A, B, or C are not met, the plant must be brought to a MODE in which overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours.
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| 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. 13) 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.
| |
| Required Action D.1 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 3. This Note prohibits the use of LCO 3.0.4.a to enter MODE 3 during startup with the LCO not met.
| |
| However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 3, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
| |
| 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)
| |
| GRAND GULF B 3.5-7 LBDCR 14043
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| | |
| ECCS ² Operating B 3.5.1 BASES ACTIONS E.1 (continued)
| |
| The LCO requires eight ADS valves to be OPERABLE to provide the ADS function. Reference 14 contains the results of an analysis that evaluated the effect of one ADS valve being out of service. Per this analysis, operation of only seven ADS valves will provide the required depressurization. However, 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 (Ref. 12) and has been found to be acceptable through operating experience. < 7 F.1 and F.2 -27)68
| |
| If any one low pressure ECCS injection/spray subsystem is inoperable in addition to one inoperable ADS valve, adequate core cooling is ensured by the OPERABILITY of HPCS and the remaining low pressure ECCS injection/spray subsystems. However, the overall ECCS reliability is reduced because a single active component failure concurrent with a design basis LOCA could result in the minimum required ECCS equipment not being available. Since both a portion of a high pressure (ADS) and a low pressure subsystem are inoperable, a more restrictive Completion Time of 72 hours is required to restore either the low pressure ECCS injection/spray subsystem or the ADS valve to OPERABLE status.
| |
| This Completion Time is based on a reliability study (Ref. 12) and has been found to be acceptable through operating experience.
| |
| < 7 G.1 -27)68
| |
| If any Required Action and associated Completion Time of Condition E or F are not met or if two or more ADS valves are inoperable, the plant must be brought to a MODE in which overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within12 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. 13) 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)
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| GRAND GULF B 3.5-8 LBDCR 14043
| |
| | |
| RCIC System B 3.5.3 BASES ACTIONS Applicability with the LCO not met after performance of a (continued) risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
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| A.land A,2 -27)68
| |
| If the RCIC System is inoperab l e during MODE l, or MODES 2 or 3 with reactor steam dome pressure> 150 sig, and the HPCS System is verified to be OPERABLE, the CIC System must be restored to OPERABLE status within 14 day In this Condition, loss of the RCIC System will not affect the overall plant capability to provide makeup inventory at high GRAND GULF B 3.5-22a LDC ~
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| | |
| I Primary Containment Air Locks B 3.6.1.2 BASES ACTIONS B.l, B.2, and B.3 (continued) these doors to be verified locked closed by use of administrative controls. Allowing verification by administrative controls is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of the door, once it has been verified to be in the proper position, is small.
| |
| C.l, C.2, and C.3 With one or more air locks inoperable for reasons other than those described in Condition A or B, Required Action C.l 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 both doors in an air lock have failed a seal test or if the overall air lock leakage is not within limits. In many instances (e.g., only one seal per door has failed) 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 affected primary containment air locks must be verified closed. This Required 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 OP st tus within 1 hour.
| |
| -27)68
| |
| Additionally, the air e res ore o PERABLE status within 24 hour. The 24 hour Completion Time is reasonable for restoring an inoperable air lock to OPERABLE status considering that at least one door is maintained closed in each affected air lock.
| |
| D.l 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 (continued)
| |
| GRAND GULF B 3.6-10 Revision No. -e-
| |
| | |
| PCIVs B 3.6.1.3 BASES ACTIONS A.land A.2 (continued) rating exceeds the containment design pressure rating. This preserves both the containment isolation function and the system overpressure protection function. The Note also avoids unnecessary safety system unavailability time and unnecessary occupational dose that would be associated with de-activating the relief valve. The Note applies to relief valves employed as isolation devices in either the backflow or forward (relief) flow direction. The failure of a relief valve to remain closed during or following an accident is considered a low probability because relief valves are passive isolation devices that do not require mechanical movement to perform the isolation function and the relief setpoint provides sufficient margin to preclude the potential for premature opening due to containment post-accident pressures. Relief valves that are one-inch or smaller provide an additional physical barrier because the size restriction would limit leakage such that a large early release would not occur. Penetration configurations that meet Criterion 2 provide an additional physical barrier of a closed system. In the unlikely event that a relief valve larger than one-inch were to fail to remain closed, the leakage would be into a system which forms a closed loop outside primary containment and any containment leakage would return to primary containment through this closed loop. In accordance with Reference 4, a closed system outside the containment shall meet Quality Group Band Seismic Category 1 standards. Valves which isolate the branch lines of these closed systems are normally closed and under strict administrative control. Typical closed systems used as isolation barriers are identified in Tables 6.2-44 and 6.2-49 of Reference 2. For penetrations isolated in accordance with Required Action A.1, the device used to isolate the penetration should be the closest one available
| |
| -27)68 to the primary containment. The Required Action must be comp within the 4 hour Completion Time (8 hours for main steam 11 The specified time period of 4 hours is 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 allows a period of time to restore the MSIVs to OPERABLE status given the fact that MSIV closure will result in isolation of the main steam line(s) and a potential fo plant shutdown.
| |
| -27)68
| |
| con inued GRAND GULF B 3.6-17a LDC 87837
| |
| | |
| PCIVs B 3.6.1.3 BASES ACTIONS A.land A.2 (continued)
| |
| For affected penetrations that have been isolated in
| |
| , accordance .with Required Action A.l, the affected penetration flow path 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, and no longer capable of being automatically isolated, will be isolated should an event occur. This Required Action does not require any testing or device manipulation. Rather, it involves verification that those devices outside primary containment, drywell, and steam tunnel and capable of being mispositioned are in .,.i..i.....__ _ _ _ _ __,
| |
| correct position. The Completion Time for JSPPS[MRK MWSPEXMSR verification of "once per 31 days isolation devices outside primary containment, drywell, and steam tunnel," is appropriate because the devices are operated under administrative controls and the probability of their misalignment is low. For devices inside primary containment, drywell, or steam tunnel, the specified time period of "prior to entering MODE 2 or 3 from 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.
| |
| Required Action A.2 is modified by a Note that 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. Therefore, the probability of misalignment; once they have been verified to be in the proper position, is low .
| |
| .!L.l With one or more penetration flow paths with two PCIVs inoperable except due to leakage not within lim i ts, either the inoperable PCIVs must be restored to OPERABLE status or the affected penetration flow path must be isolated within 1 hour. The method of isolation must include the use of at least one isolation barrier that cannot be adversely continued GRAND GULF B 3.6-18 LDC 07037
| |
| | |
| PCIVs B 3.6.1.3 BASES ACTIONS C.l (continued)
| |
| With the hydrostatic leakage rate or MSIV leakage rate not within limit, the assumptions of the safety analysis may not be met. Therefore, the leakage must be restor~d to within limit within 4 hours. Restoration can be accomplished by isolating the penetration that caused the limit to be exceeded by use of one closed and de-activated automatic valve, closed manual valve, or blind flange. When a penetration is isolated, the leakage rate for the isolation penetration is assumed to be the actual pathway leakage through the isolation device. If two isolation devices are used to isolate the penetration, the leakage rate is assumed to be the lesser actual pathway leakage of the two devices.
| |
| The 4 hour Completion Time is reasonable considering the time required to restore the leakage by isolating the penetration and the relative importance to the overall containment function.
| |
| D.l, D.2, and D.3 In the event one or more primary containment purge valves are not within the purge valve leakage limits, purge valve leakage must be restored to within limits or the affected penetration must be isolated. The method of isolation must be by 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, closed manual valve, and blind flange. If a purge valve with resilient seals is utilized to satisfy Required Action D.l it must have been demonstrated to meet the leakage requirements of SR 3.6.1.3.5. The specified Completion Time is reasonable, considering that one primary containment purge valve remains closed, so that a gross breach of primary containment does not exist. ~ ______, -27)68 I In accordance with Required Action D.2, this penetration flow pa must be verifi~d to be isolated on a periodic basi he periodic verification is necessary to ensure JSPPS[MRK MWSPEXMSR t primary containment penetrations required to be isolated following an accident, which are no longer capable of being automatically isolated, will be isolated should an event occur. This Required Action does not require any testing or valve manipulation. Rather, it involves (continued)
| |
| GRAND GULF B 3.6-19 Revision No. +
| |
| | |
| PCIVs B 3.6.1.3 BASES ACTIONS D.1, D.2, and D.3 (continued) verification that those isolation devices outside primary containment and potentially capable of being mispositioned are in the correct position. For the isolation devices inside primary containment, the time period specified as "prior to entering MODE 2 or 3, from MODE 4 if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of administrative controls that will ensure that isolation device misalignment is an unlikely possibility.
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| For the primary containment purge valve with resilient seal that is isolated in accordance with Required Action D.1, SR 3.6.1.3.5 must be performed at least once every 92 days. This provides assurance that degradation of the resilient seal is detected and confirms that the leakage rate of the primary containment purge valve does not increase during the time the penetration is isolated. Since more reliance is placed on a single valve while in this Condition, it is prudent to perform the SR more often. Therefore, a Frequency of once per 92 days was chosen and has been shown acceptable based on operating experience.
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| JSPPS[MRK MWSPEXMSR E.1 and E.2 If any Required Action and associated Completion Time cannot be met in MODE 1, 2, or 3, 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.
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| F.1 If any Required Action and associated Completion Time cannot be met, the plant must be placed in a condition in which the LCO does not apply. If applicable, movement of recently irradiated fuel assemblies in the primary and (continued)
| |
| GRAND GULF B 3.6-20 LBDCR 18128
| |
| | |
| LLS Valves B 3.6.1.6 BASES APPLICABILITY In MODES 1, 2, and 3, an event could cause pressurization of the reactor and opening of S/RVs. 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 LLS valves OPERABLE is not required in MODE 4 or 5.
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| ACTIONS A.1 With one LLS valve inoperable, the remaining OPERABLE LLS valves are 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 LLS S/RVs and the low probability of an event in which the remaining LLS S/RV capability would be inadequate.
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| -27)68
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| B.1 If the inoperable LLS valve cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours.
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| 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 to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.
| |
| Required Action B.1 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 3. This Note prohibits the use of LCO 3.0.4.a to enter MODE 3 during startup with the LCO not met.
| |
| However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 3, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
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| 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)
| |
| GRAND GULF B 3.6-33 LBDCR 14043
| |
| | |
| RHR Containment Spray System B 3.6.1.7 BASES APPLICABLE The containment spray operation is also assumed in the SAFETY ANALYSES design basis LOCA dose analysis to scrub iodine from the (continued) containment atmosphere thereby mitigating the affects of the accident.
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| The RHR Containment Spray System satisfies Criterion 3 of the NRC Policy Statement.
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| LCO In the event of a Design Basis Accident (DBA), a minimum of one RHR containment spray subsystem is required to mitigate potential bypass leakage paths and maintain the primary containment peak pressure below design limits. To ensure that these requirements are met, two RHR containment spray subsystems must be OPERABLE. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. An RHR containment spray subsystem is OPERABLE when the pump, the heat exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE.
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| Management of gas voids is important to RHR Containment Spray System OPERABILITY.
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| APPLICABILITY In MODES 1, 2, and 3, a DBA could cause 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, maintaining RHR containment spray subsystems OPERABLE is not required in MODE 4 or 5.
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| ACTIONS A.1 -27)68
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| J With one RHR containment spray subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE RHR containment spray 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 was chosen in light of the redundant RHR containment capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
| |
| (continued)
| |
| GRAND GULF B 3.6-37 LBDCR 16003
| |
| | |
| RHR Suppression Pool Cooling B 3.6.2.3 BASES APPLICABLE The RHR Suppression Pool Cooling System satisfies SAFETY ANALYSES Criterion 3 of the NRC Policy Statement.
| |
| (continued)
| |
| 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 the design limits (Ref. 1). To ensure that these requirements are met, two RHR suppression pool cooling subsystems must be OPERABLE. 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 OPERABLE when the pump, two heat exchangers, 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 a release of radioactive material to primary containment and cause 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
| |
| -27)68
| |
| 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 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 cooling capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
| |
| B.1 If one RHR suppression pool cooling subsystems inoperable and is not restored to OPERABLE status within the required Completion Time, the plant must be brought to a condition in which overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours.
| |
| (continued)
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| GRAND GULF B 3.6-57 LBDCR 14044
| |
| | |
| SPMU System B 3.6.2.4 BASES ACTIONS B.1 (continued)
| |
| Therefore, the upper containment pool water temperature must be restored to within limit within 24 hours. The 24 hour Completion Time is sufficient to restore the upper containment pool to within the specified temperature limit. It also takes into account the low probability of an event occurring that would require the SPMU System.
| |
| C.1
| |
| -27)68
| |
| With one SPMU subsystem inoperable for reasons other than Condition A or B, the inoperable subsystem must be restored to OPERABLE status within 7 days. The 7 day Completion Time is acceptable in light of the redundant SPMU System capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
| |
| D.1 and D.2 If any Required Action and required Completion Time cannot 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.
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| SURVEILLANCE SR 3.6.2.4.1 REQUIREMENTS The upper containment pool water level is regularly monitored to ensure that the required limits are satisfied. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
| |
| (continued)
| |
| GRAND GULF B 3.6-63 LBDCR 18127
| |
| | |
| Drywell Air Lock B 3.6.5.2 BASES ACTIONS 8.1, 8.2, and B.3 (continued) drywell under the control of a dedicated individual stationed at the air lock to ensure that only one door is opened at a time.
| |
| C.l and C.2 With tne air lock inoperable for reasons other than those described in Condition A or B, Required Action C.l requires that one door in the drywell air lock must be verified to be closed. This Required Action must be completed within the 1 hour Completion Time. This specified time period is consistent with the ACTIONS of LCO 3.6.5.1, which requires that the drywell be restored to OPERABLE status within 1 hour.
| |
| Additionally, the air lock must be restored to OPERABLE status within 24 hour. The 24 hour Completion Time is reasonable for resto g an inoperable air lock to OPERABLE status, considering hat at least one door is maintained closed in the air 1 k.
| |
| -27)68
| |
| D.1 and 0.2 If the inoperable drywell air lock cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MOOE 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, (continued)
| |
| GRAND GULF 8 3.6-110 Revision No. r
| |
| | |
| Drywell Isolation Valve(s)
| |
| B 3.6.5.3 BASES ACTIONS A.I and A.2 (continued)
| |
| With one or more penetration flow paths with one drywell isolation valve inoperable, the affected penetration flow path 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. In this condition, the remaining OPERABLE drywell isolation valve is adequate to perform the isolation function. However, the overall reliability is reduced because a single failure in the OPERABLE drywell isolation valve could result in a loss of drywell isolation. The 8 hour Completion Time is acceptable, due to the low probability of the inoperable valve resulting in excessive drywell leakage and the low probability of the limiting event for drywell leakage occurring during this short time frame. In addition, the Completion Time is reasonable, considering the time required to isolate the penetration and the relative importance of supporting drywell OPERABILITY during MODES 1, 2, and 3. N -27)68 I For affected penetration flow paths that have been isolated in accordance with Required Action A.I, the affected penetrations must be verified to be isolated on a periodic basis. This is necessary to ensure that drywell penetrations that are required to be isolated following an accident, and are no longer capable of being automatically isolated, will be isolated should an event occur. This
| |
| * Required Action does not require any testing or device manipulation; rather, it involves verification that those devices outside drywell and capable of potentially being mispositioned are in the correct position. Since these devices are inside primary containment, the time period specified as "prior to entering MODE 2 or 3 from 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 other administrative controls that will ensure that misalignment is an unlikely possibility. Also, this Completion Time is consistent with (continued)
| |
| GRAND GULF B 3.6-116 Revision No. +-
| |
| | |
| SSW System and UHS B 3.7.1 BASES APPLICABILITY In MODES 1, 2, and 3, the SSW System and the UHS are required to be OPERABLE to support OPERABILITY of the equipment serviced by the SSW System and UHS and required to be OPERABLE in these MODES.
| |
| In MODES 4 and 5, the OPERABILITY requirements of the SSW System and UHS are determined by the systems they support.
| |
| ACTIONS A.1 If one cooling tower has one fan inoperable, action must be taken to restore the inoperable cooling tower fan to OPERABLE status within 7 days.
| |
| -27)68 The 7 day Completion Time is reasonable, based on the low probability of an accident occurring during the 7 days that one cooling tower fan is inoperable, the number of available systems, and the time required to complete the Required Action.
| |
| B.1 and D.1 -27)68
| |
| If one SSW subsystem is inoperable or if both fans in one cooling tower are inoperable (since this is equivalent to the loss of function of one SSW subsystem), it must be restored to OPERABLE status within 72 hours.
| |
| With the unit in this condition, the remaining OPERABLE SSW subsystem is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure in the OPERABLE SSW subsystem could result in loss of SSW function. The 72 hour Completion Time was developed taking into account the redundant capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
| |
| The Required Action is modified by two Notes indicating that the applicable Conditions of LCO 3.8.1, "AC Sources - Operating," and LCO 3.4.9, "Residual Heat Removal (RHR) Shutdown Cooling System -
| |
| Hot Shutdown," be entered and the Required Actions taken if the inoperable SSW subsystem results in an inoperable DG or RHR shutdown cooling subsystem, respectively. This is in accordance with LCO 3.0.6 and ensures the proper actions are taken for these components.
| |
| (continued)
| |
| GRAND GULF B 3.7-4 Revision No. 0
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| | |
| AC Sources -Operating B 3 .8.1 BASES (continued)
| |
| ACTIONS A Note prohibits the application of LCO 3.0.4 . b to an inoperable DG. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable DG 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.
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| A.....l.
| |
| To ensure a highly reliable power source remains, it is necessary to verify the availability of the remaining required offsite circuits on a more frequent basis. Since the Required Action only specifies "perform," a failure of SR 3.8.1.1 acceptance criteria does not result in the Required Action not met. However, if a second required circuit fails SR 3.8.1.1, the second offsite circuit is inoperable, and Condition C, for two offsite circuits inoperable, is entered.
| |
| Ll According to Regulatory Guide 1.93 (Ref. 6), operation may continue in Condition A for a period that should not exceed 72 hours.
| |
| -27)68
| |
| This Completion Time assumes sufficient offsite power remains to power the minimum loads needed to respond to analyzed events. In the event more than one division is without offsite power, this assumption is not met.
| |
| Therefore, the optional Completion Time is specified.
| |
| Should two (or more) divisions be affected, the 24 hour Regulatory Guide assumptions supporting a 24 hour
| |
| ~I---------.
| |
| Completion Time is conservative with respect to the -27)68
| |
| J Completion Time for both offsite circuits inoperable. ~ Wit h 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 lE distribution system.
| |
| GRAND GULF B 3.8 - 5 LDC 07028
| |
| | |
| AC Sources Operating B 3.8.1 BASES ACTIONS B.3.1 and B.3.2 (continued)
| |
| Required Action B.3.1 provides an allowance to avoid unnecessary testing of OPERABLE DGs. If it can be determined that the cause of the inoperable DG does not exist on the OPERABLE DG, SR 3.8.1.2 does not have to be performed. If the cause of inoperability exists on other DG(s), the other DG(s) are declared inoperable upon discovery, and Condition E and potentially Condition H of LCO 3.8.1 is entered. Once the failure is repaired, and the common cause failure no longer exists, Required Action B.3.1 is satisfied. If the cause of the initial inoperable DG cannot be confirmed not to exist on the remaining DG(s),
| |
| performance of SR 3.8.1.2 suffices to provide assurance of continued OPERABILITY of those DG(s).
| |
| In the event the inoperable DG is restored to OPERABLE status prior to completing either B.3.1 or B.3.2, the 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.
| |
| According to Generic Letter 84-15 (Ref. 7), 24 hours is 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. Although Condition B applies to a single inoperable DG, several Completion Times are specified for this Condition.
| |
| Division 3 DG
| |
| -27)68
| |
| The first Completion Time applies to an inoperable Division 3 DG. The 72 hour Completion Time takes into account the capacity and capability of the remaining AC sources, reasonable time for repairs, and low probability of a DBA occurring during this period. This Completion Time begins only upon discovery of an inoperable Division 3 DG and, as such, provides an exception to the normal time zero for beginning the allowed outage time clock (i.e., for beginning the clock (continued)
| |
| GRAND GULF B 3.8-8 LBDCR 16007
| |
| | |
| AC Sources Operating B 3.8.1 BASES ACTIONS B.4 (continued)
| |
| : 5. The RCIC high pressure injection system and the Division 1 and 2 DGs will not be taken out of service for planned maintenance while the Division 3 DG is out of service during the extended allowed outage time.
| |
| -27)68
| |
| Division 1 and 2 The second Completion Time (14 days) applies to an inoperable Division 1 or Division 2 DG and is a risk-informed allowed outage time (AOT) based on a plant specific risk analysis. The extended AOT would typically be used for voluntary planned maintenance or inspections but can also be used for corrective maintenance.
| |
| However, use of the extended AOT for voluntary planned maintenance should be limited to once within an operating cycle (24 months) for each DG (Division 1 and Division 2).
| |
| Additional contingencies are to be in place for any extended AOT duration (greater than 72 hours and up to 14 days) as follows:
| |
| : 1. Weather conditions will be evaluated prior to entering an extended DG AOT for voluntary planned maintenance. An extended DG AOT will not be entered for voluntary planned maintenance purposes if official weather forecasts are predicting severe conditions (hurricane, tropical storm, tornado, or snow/ice storm) that could significantly threaten grid stability during the planned outage time.
| |
| : 2. The condition of the offsite power supply and switchyard will be evaluated.
| |
| : 3. No elective maintenance will be scheduled within the switchyard that would challenge offsite power availability during the proposed extended DG AOT.
| |
| : 4. Operating crews will be briefed on the DG work plan whenever the extended AOT period is used, with consideration given to key procedural actions that would be required in the event of a loss of offsite power or station blackout. It is expected that the Division 3 DG can be cross-connected and ready to power required shutdown equipment on either Division 1 or Division 2 ESF bus within two hours of determining a need to cross-connect.
| |
| (continued)
| |
| GRAND GULF B 3.8-8b LBDCR 16007
| |
| | |
| AC Sources-Operating B 3.8.1 BASES ACTIONS C.l and C.2 (continued)
| |
| With both of the required offsite circuits inoperable, sufficient onsite AC sources are available to maintain the unit in a safe shutdown condition in the event of a OBA or 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. 6), 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 offsite source is restored within 24 hours, power operation continues in accordance with Condition A.
| |
| -27)68
| |
| D.1 and D.2 Pursuant to LCO 3.0.6, the Distribution System ACTIONS ~uld not be entered even if all AC sources to it were inoperable, resulting in de-energization. Therefore, the Required Actions of Condition Dare modified by a Note to indicate that when Condition Dis entered with no AC source to any division, Actions for LCO 3.8.7, "Distribution Systems-Operating," must be immediately entered. This allows Condition D to provide requirements for the loss of the offsite circuit and one 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. 6), 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 (continued)
| |
| GRAND GULF B 3.8-11 Revision No. -&-
| |
| | |
| AC Sources-Operating B 3.8.1 BASES ACTIONS D.l and D.2 {continued) is offset by the susceptibility of this power system configuration to a single bus or switching failure. The 12 hour Completion Time takes into account the capacity and capability of the remaining AC sources, reasonable time for repairs, and low probability of a OBA occurring during this period.
| |
| -27)68
| |
| E.l With two DGs inoperable, there is one rema1n1ng standby AC source. Thus, with an assumed loss of offsite electrical power, insufficient standby AC sources are 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 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. 6), with both DGs inoperable, operation may continue for a period that should not exceed 2 hours. This Completion Time assumes complete loss of onsite {DG) AC capability to power the minimum loads needed to respond to analyzed events. In the event Division 3 DG in conjunction ~ith Division 1 or 2 DG is inoperable, with Division 1 or 2 remaining, a significant spectrum of breaks would be capable of being responded to with onsite power. Even the worst case event would be mitigated to some extent-an extent greater than a typical two division design in which this condition represents complete loss of onsite power function. Given the remaining funr.tion. a 24 hour Comoletion Time is aoorooriate. At the
| |
| ~~~-~i-i~i; ~~~~~~~;~~~~; - ~~vision 3 s~~te~s could be declared inoperable {see Applicability Note) and this Condition could be exited with only one required DG (continued)
| |
| GRAND GULF B 3.8-12 Revision No. -G-
| |
| | |
| AC Sources ² Operating B 3.8.1 BASES ACTIONS E.1 (continued) remaining inoperable. However, with a Division 1 or 2 DG remaining inoperable and the HPCS declared inoperable, a redundant required feature failure exists, according to Required Action B.2.
| |
| F.1 Each sequencer is an essential support system to both the offsite circuit and the DG associated with a given ESF bus. Furthermore, the sequencer(s) is on the primary success path for most major AC electrically powered safety systems powered from the associated ESF bus. Although loss of an ESF bus's sequencer potentially affects the major ESF systems in the division, a design basis event with the worst single failure would not result in a complete loss of onsite power function (DGs) and would be mitigated to some extent by the redundant onsite sources. In addition, operator action to start the DG affected by the inoperable sequencer and manually connect the required ESF loads to either the affected DG or an available offsite source represents a significant benefit justifying an extended Completion Time over the condition of one DG and one offsite circuit inoperable. The 24 hour Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining sequencer OPERABILITY. This time period also ensures that the probability of an accident requiring sequencer OPERABILITY occurring during periods when the sequencer is inoperable is minimal.
| |
| G.1 -27)68
| |
| _ _ _____.I 1 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 overall plant risk is minimized. To achieve this status, the unit must be brought to 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. 8) 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.
| |
| Required Action G.1 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 3. This Note prohibits the use of LCO 3.0.4.a to enter MODE 3 during startup with the LCO not met.
| |
| However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment (continued)
| |
| GRAND GULF B 3.8-13 LBDCR 14043
| |
| | |
| DC Sources-Operating B 3.8.4 BASES ACTIONS A.l (continued) imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for comp1ete loss of DC power to the affected division . The additional time provided by the Completion Time is consistent with the capability of the battery to maintain its short term capability to respond to a design basis event.
| |
| A Note is added to take exception to the allowance of LCO 3.0.4 to enter MODES or other specified conditions in the Applicability. Even though Condition A Required Actions do not in themselves require a plant shutdown, or require exiting the MODES or other specified conditions in the Applicability, the condition of the DC system is not such that extended operation is expected. Therefore, the Note would require restoration of an*inoperable battery charger to OPERABLE status prior to increasing power. This exception is not intended to preclude the allowance of LCO 3.0.4 to always enter MODES or other specified conditions in the Applicability as a result of a plant shutdown.
| |
| 8.1 If the battery cell parameters cannot be maintained within the Category A limits, the short term capability of the battery is also degraded and the battery must be declared inoperable.
| |
| C. l Condition C represents one division 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 DC power to the affected division. The 2 hour limit is consistent with the allowed time for an inoperable DC distribution system division.
| |
| -27)68
| |
| (continued)
| |
| GRAND GULF 8 3.8-54 Revision No. +
| |
| | |
| Distribution Systems-Operating 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 OBA.
| |
| A Note has been added taking exception to the Applicability requirements for the Division 3 electric power distribution subsystem, provided the HPCS System is declared inoperable.
| |
| This exception is intended to allow declaring of HPCS inoperable either in lieu of declaring the Division 3 electric power distribution subsystem inoperable, or at any time subsequent to entering ACTIONS for an inoperable Division 3 electric power distribution subsystem. This exception is acceptable since, with HPCS inoperable and the associated ACTIONS entered, the Division 3 AC electric power
| |
| ,.a.:,..+-..:1-..,,.f..,:,,._ ,..,,&-...,..11 .... +,..,..,. ....,__,,.".,:,4,,.,.. '"'""' .... AA.:+.,:,-..,...,.1 ""lrr l\l'l""l,ftr- n+
| |
| Ul~l,I IUUl,IUII ~UU~J~l.t:111 fJIUVIUt:~ IIU QUUll,IUIIQI Q~~UIQll\..t: UI meeting the above criteria.
| |
| Electrical power distribution subsystem requirements for MODES 4 and 5 are covered in the Bases for LCO 3.8.8, "Di stri but ion Systems-Shutdown." I ACTIONS A.1 With one or more Division 1 or 2 required AC buses, load centers, motor control centers, or distribution panels in one division inoperable, 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 power distribution subsystems could result in the minimum required ESF functions not being supported. Therefore, the required AC buses, load centers, motor control centers, and distribution panels must be restored to OPERABLE status within 8 ho' J
| |
| -27)68
| |
| (continued)
| |
| GRAND GULF B 3 . 8-74 Revision No.
| |
| Distribution Systems Operating B 3.8.7 BASES ACTIONS B.1 With one or more Division 1 or 2 DC electrical power distribution subsystems inoperable, the remaining DC 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 DC electrical power distribution subsystems could result in the minimum required ESF functions not being supported. Therefore, the required DC buses must be restored to OPERABLE status within 2 hours by powering the bus from the associated battery or charger.
| |
| ~--------
| |
| -27)68
| |
| Condition B may represent one division without adequate DC power, potentially with both the battery significantly 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 divisions, and restoring power to the affected division.
| |
| This 2 hour limit is more conservative than Completion Times allowed for the majority of components that could be without power. Taking exception to LCO 3.0.2 for components without adequate DC power, that 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 (continued)
| |
| GRAND GULF B 3.8-76 LBDCR 20084
| |
| | |
| Attachment 4 GNRO-2023/00014 Cross-Reference of TSTF-505 and GGNS Technical Specifications (18 pages follow)
| |
| | |
| GNRO-2023/00014 Attachment 4 Page 1 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
| |
| Completion Times 1.3 1.3 Example 1.3-8 Example 1.3-8 Example 1.3-8 N/A TSTF-505 changes are incorporated.
| |
| The proposed example varies from the TSTF to gain consistency with the GGNS TS. Specifically, the Required Action and Completion Time in the proposed Example 1.3-8 will be to place the unit in MODE 3 in 12 hours (versus in MODE 3 in 6 hours) and MODE 4 in 36 hours (versus in MODE 5 in 36 hours).
| |
| Standby Liquid Control (SLC) 3.1.7 3.1.7 System One SLC subsystem inoperable 3.1.7.B.1 3.1.7.D.1 Yes TSTF-505 changes are incorporated.
| |
| for other reasons.
| |
| In this Condition, the remaining OPERABLE subsystem is adequate to perform the shutdown function.
| |
| Note - A loss of function (LOF) is a loss of safety function as defined per the Safety Function Determination Program, in accordance with TS 5.5.10. A Condition that has the potential to represent a LOF should include a note to preclude application of a RICT during a LOF. The concern is inappropriate extension of Completion Times (CTs), via a RICT, during a LOF situation. However, a note to preclude a RICT during a LOF is unnecessary when there are other related Required Actions that are specific to LOF or loss of trip capability situations. The CTs for the Actions that are associated with a TS LOF, or a loss of trip capability, are generally very short (e.g., one or two hours); therefore, these short CTs alleviate the concern of inappropriate extension of CTs during a LOF making a note to preclude RICT entry unnecessary. Where appropriate, these issues are addressed in the "Comments" column of Table A4-1.
| |
| | |
| GNRO-2023/00014 Page 2 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
| |
| Reactor Protection System 3.3.1.1 3.3.1.1 (RPS) Instrumentation One or more required channels 3.3.1.1.A.1 3.3.1.1.A.1 Yes TSTF-505 changes are incorporated.
| |
| inoperable. 3.3.1.1.A.2 3.3.1.1.A.2 Yes A note to preclude RICT entry during a TS LOF is not necessary since Condition C addresses one or more Functions with RPS trip capability not maintained and trip capability must be restored within one hour. A RICT is not proposed for Condition C.
| |
| One or more Functions with one 3.3.1.1.B.1 3.3.1.1.B.1 Yes TSTF-505 changes are incorporated.
| |
| or more required channels 3.3.1.1.B.2 3.3.1.1.B.2 Yes inoperable in both trip systems.
| |
| See above. Condition C addresses a TS LOF and has a one-hour Action statement; therefore, a note to preclude RICT entry during a TS LOF is unnecessary for Condition B Actions.
| |
| Source Range Monitor (SRM) 3.3.1.2 3.3.1.2 Instrumentation One or more required SRMs 3.3.1.2.A.1 3.3.1.2.A.1 No TSTF-505 changes are not incorporated.
| |
| inoperable in MODE 2 with intermediate range monitors Entergy is not proposing a RICT for this Condition.
| |
| (IRMs) on Range 2 or below.
| |
| | |
| GNRO-2023/00014 Page 3 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
| |
| End of Cycle Recirculation 3.3.4.1 3.3.4.1 Pump Trip (EOC-RPT)
| |
| Instrumentation One or more required channels 3.3.4.1.A.1 3.3.4.1.A.1 Yes TSTF-505 changes are incorporated.
| |
| inoperable. 3.3.4.1.A.2 3.3.4.1.A.2 Yes Condition A addresses one or more channels inoperable, but with the EOC-RPT trip capability maintained. Condition B addresses a loss of trip capability and requires restoration, or applies operational limitations, within two hours. A RICT is not proposed for Condition B. Therefore, it is unnecessary to apply a note in Condition A to preclude RICT entry when trip capability is not maintained.
| |
| Anticipated Transient Without 3.3.4.2 3.3.4.2 Scram Recirculation Pump Trip (ATWS-RPT)
| |
| Instrumentation One or more channels 3.3.4.2.A.1 3.3.4.2.A.1 Yes TSTF-505 changes are incorporated.
| |
| inoperable. 3.3.4.2.A.2 3.3.4.2.A.2 Yes Condition A addresses one or more channels inoperable, but with ATWS-RPT trip capability maintained. Condition B addresses a loss of trip capability within one Function and Condition C addresses a loss of trip capability for both Functions. A RICT is not proposed for Conditions B or C. Condition C is very limiting for a loss of trip capability for both Functions (one hour CT).
| |
| Therefore, it is unnecessary to apply a note to Condition A CTs to preclude RICT entry when trip capability is not maintained.
| |
| | |
| GNRO-2023/00014 Page 4 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
| |
| Emergency Core Cooling 3.3.5.1 3.3.5.1 System (ECCS)
| |
| Instrumentation As required by Required Action 3.3.5.1.B.3 3.3.5.1.B.3 Yes TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table 3.3.5.1-1. Action B.1 addresses a loss of low pressure ECCS automatic initiation capability and Action B.2 addresses a loss of High-Pressure Core Spray (HPCS) automatic initiation capability. Both require supported features be declared inoperable within one hour from discovery of loss of initiation capability. A note to preclude RICT entry during a LOF is not necessary for Action B.3 as a loss of capability will result in declaring the supported system inoperable within one hour.
| |
| As required by Required Action 3.3.5.1.C.2 3.3.5.1.C.2 Yes TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table 3.3.5.1-1. Action C.1 addresses a loss of low pressure ECCS automatic initiation capability and has a one-hour CT. A note to preclude RICT entry during a TS LOF is not necessary for Action C.2 as a loss of automatic initiation capability is addressed by C.1 t
| |
| As required by Required Action 3.3.5.1.D.2.1 3.3.5.1.D.2.1 Yes TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table 3.3.5.1-1. A note to preclude RICT entry during a TS LOF is not necessary for Action D.2.1 as Action D.1 addresses a loss of HPCS System automatic suction swap and has a one-hour CT. Note that if the HPCS pump is aligned to the suppression pool, the Function is already performed.
| |
| | |
| GNRO-2023/00014 Page 5 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
| |
| As required by Required Action 3.3.5.1.E.2 3.3.5.1.E.2 Yes TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table 3.3.5.1-1. Action E.1 addresses a loss of automatic initiation capability for low pressure ECCS Functions and has a one-hour CT. However, Action E.1 is not applicable to HPCS Functions 3.f and 3.g since the loss of one channel results in a loss of the Function (one-out-of-one logic). This loss is considered acceptable during the seven days allowed by Action E.2. Therefore, its acceptable to apply a RICT to Action E.2 except during a LOF (e.g., for ECCS). A note to preclude RICT entry during a LOF is included for Action E.2.
| |
| As required by Required Action 3.3.5.1.F.2 3.3.5.1.F.2 Yes TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table 3.3.5.1-1. A note to preclude RICT entry during a TS LOF is not necessary for Action F.2 as a loss of initiation capability for the Automatic Depressurization System (ADS) requires Action per F.1 and has a one-hour CT.
| |
| As required by Required Action 3.3.5.1.G.2 3.3.5.1.G.2 Yes TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table 3.3.5.1-1. A loss of ADS initiation capability requires Action per G.1, which has a one-hour CT. Action G.1 is not applicable to Functions 4.h and 5.g since they are Manual Initiation Functions and are not assumed in any accident or transient analysis. For these reasons, a note to preclude RICT entry during a TS LOF is not necessary for Action G.2.
| |
| | |
| GNRO-2023/00014 Page 6 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
| |
| Reactor Core Isolation Cooling 3.3.5.2 3.3.5.3 (RCIC) System Instrumentation As required by Required Action 3.3.5.2.B.2 3.3.5.3.B.2 Yes TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table 3.3.5.2-1. Action B.1 addresses a loss of RCIC System initiation capability and requires RCIC be declared inoperable within one hour from discovery. Also, a loss of RCIC is not a TS LOF as its not assumed to function in any accident or transient analyses.
| |
| As required by Required Action 3.3.5.2.D.2.1 3.3.5.3.D.2.1 Yes TSTF-505 changes are incorporated.
| |
| A.1 and referenced in Table 3.3.5.2-1. Action D.1 addresses a loss of RCIC System automatic suction swap and requires RCIC be declared inoperable within one hour from discovery.
| |
| Also, a loss of RCIC is not a TS LOF as its not assumed to function in any accident or transient analyses.
| |
| | |
| GNRO-2023/00014 Page 7 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
| |
| Primary Containment Isolation 3.3.6.1 3.3.6.1 GGNS TS is titled Primary Containment and Instrumentation Drywell Isolation Instrumentation.
| |
| One or more required channels 3.3.6.1.A.1 3.3.6.1.A.1 Yes TSTF-505 changes are incorporated.
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| inoperable.
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| A note to preclude RICT entry during a TS LOF is not necessary for Action A.1 as Condition B would be entered if multiple, inoperable, untripped channels within the same Function result in redundant automatic isolation capability being lost for the associated penetration flow path(s). A RICT is not applied to the Condition B CT and isolation capability must be restored within one hour.
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| Residual Heat Removal (RHR) 3.3.6.3 3.3.6.3 Containment Spray System Instrumentation As required by Required Action 3.3.6.3.B.2 3.3.6.3.B.2 Yes TSTF-505 changes are incorporated.
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| A.1 and referenced in Table 3.3.6.3-1 A note to preclude RICT entry during a TS LOF is not necessary for Action B.2 as Action B.1 addresses a loss of RHR Containment Spray initiation capability in both trip systems and requires declaring the associated subsystems inoperable within one hour from discovery. A RICT is not applied to the Condition B.1 CT.
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| GNRO-2023/00014 Page 8 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
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| As required by Required Action 3.3.6.3.C.2 3.3.6.3.C.2 Yes TSTF-505 changes are incorporated.
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| A.1 and referenced in Table 3.3.6.3-1 A note to preclude RICT entry during a TS LOF is not necessary for Action C.2 as Action C.1 addresses a loss of RHR Containment Spray initiation capability in both trip systems and requires declaring the associated subsystems inoperable within one hour from discovery. A RICT is not applied to the Condition C.1 CT.
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| Relief and Low-Low Set (LLS) 3.3.6.5 3.3.6.5 Instrumentation One trip system inoperable. 3.3.6.5.A.1 3.3.6.5.A.1 Yes TSTF-505 changes are incorporated.
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| In this Condition, the remaining OPERABLE trip system is adequate to perform the relief and LLS initiation functions.
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| Loss of Power (LOP) 3.3.8.1 3.3.8.1 Instrumentation One or more channels 3.3.8.1.A.1 3.3.8.1.A.1 Yes TSTF-505 changes are incorporated.
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| inoperable.
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| A note to preclude RICT entry during a TS LOF is not necessary for Action A.1 as Condition B addresses a loss of actuation capability, which must be restored within one hour; otherwise, the associated Diesel Generator must be declared inoperable immediately. A RICT is not applied to the Condition B CT.
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| GNRO-2023/00014 Page 9 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
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| Safety/Relief Valves (S/RVs) 3.4.4 3.4.4 One [required] S/RV inoperable. 3.4.4.A.1 3.4.4.A.1/A.2 No TSTF-505 changes are not incorporated. The Required Action of GGNS Condition A is a default condition, i.e., to perform a plant shutdown. TSTF-505 states that a RICT will not be applied to default conditions.
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| ECCS - Operating 3.5.1 3.5.1 One low pressure ECCS 3.5.1.A.1 3.5.1.A.1 Yes TSTF-505 changes are incorporated.
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| injection/spray subsystem inoperable. In this Condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA.
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| HPCS System inoperable. 3.5.1.B.2 3.5.1.B.2 Yes TSTF-505 changes are incorporated.
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| In this Condition, adequate core cooling is ensured by the OPERABILITY of the redundant and diverse low pressure ECCS injection/spray subsystems in conjunction with the ADS.
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| Two ECCS injection subsystems 3.5.1.C.1 3.5.1.C.1 Yes TSTF-505 changes are incorporated.
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| inoperable.
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| OR In this Condition, the remaining OPERABLE IOne ECCS injection and one subsystems provide adequate core cooling during a ECCS spray subsystem LOCA.
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| inoperable.
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| GNRO-2023/00014 Page 10 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
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| One ADS valve inoperable. 3.5.1.E.1 3.5.1.E.1 Yes TSTF-505 changes are incorporated.
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| The LCO requires that eight ADS valves be OPERABLE to provide the ADS Function. In this Condition, operation of seven ADS valves will provide the required depressurization.
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| One ADS valve inoperable. 3.5.1.F.1 3.5.1.F.1 Yes TSTF-505 changes are incorporated.
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| IAND 3.5.1.F.2 3.5.1.F.2 Yes One low pressure ECCS In this Condition, adequate core cooling is ensured injection/spray subsystem by the OPERABILITY of HPCS and the remaining inoperable. low pressure ECCS injection/spray subsystems.
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| RCIC System 3.5.3 3.5.3 RCIC System inoperable. 3.5.3.A.2 3.5.3.A.2 Yes TSTF-505 changes are incorporated.
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| The Function is maintained in this Condition because a loss of the RCIC System will not affect the overall plant capability to provide makeup inventory at high RPV pressure since the HPCS System is the only high-pressure system assumed to function during a LOCA.
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| GNRO-2023/00014 Page 11 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
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| Primary Containment Air 3.6.1.2 3.6.1.2 Locks One or more primary 3.6.1.2.C.3 3.6.1.2.C.3 Yes TSTF-505 changes are incorporated.
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| containment air locks inoperable In this Condition, with one or more primary for reasons other than Condition containment air locks inoperable, LCO 3.6.1.2 A or B.
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| Condition C.1 initiates action to evaluate the primary containment overall leakage rate per LCO 3.6.1.1. If the overall leakage rate exceeds limits, LCO 3.6.1.1 Condition A requires restoration of primary containment to an OPERABLE status within one hour; otherwise, a plant shutdown would be required. A RICT is not proposed for LCO 3.6.1.1 Completion Times.
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| Primary Containment Isolation 3.6.1.3 3.6.1.3 Valves (PCIVs)
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| One or more penetration flow 3.6.1.3.A.1 3.6.1.3.A.1 Yes TSTF-505 changes are incorporated.
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| paths with one PCIV inoperable except due to leakage not within There is no loss of function in this Condition limits. because one isolation valve in the flow path remains operable.
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| 3.6.1.3.A.2 3.6.1.3.A.2 N/A TSTF-505 change to add "following isolation" is incorporated.
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| GNRO-2023/00014 Page 12 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
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| One or more penetration flow 3.6.1.3.E.1 3.6.1.3.D.1 Yes TSTF-505 changes are incorporated.
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| paths with one or more containment purge valves not within purge valve leakage limits. 3.6.1.3.E.2 3.6.1.3.D.2 N/A TSTF-505 change to add "following isolation" is incorporated.
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| 3.6.1.3.E.3 3.6.1.3.D.3 N/A TSTF-505 change to add "following isolation" is incorporated.
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| Low-Low Set (LLS) Valves ------------- 3.6.1.6 One LLS valve inoperable 3.6.1.6.A.1 Yes TSTF-505 does not apply a RICT to LLS Valves; however, GGNS proposes to apply a RICT to this TS because the function can be conservatively modeled, and it meets the other criteria for inclusion.
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| In this Condition, the remaining OPERABLE LLS valves are adequate to perform the design function.
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| GNRO-2023/00014 Page 13 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
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| Residual Heat Removal (RHR) 3.6.1.7 3.6.1.7 Containment Spray System One RHR containment spray 3.6.1.7.A.1 3.6.1.7.A.1 Yes TSTF-505 changes are incorporated.
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| subsystem inoperable.
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| In this Condition, the remaining OPERABLE RHR containment spray subsystem is adequate to perform the primary containment cooling function.
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| Residual Heat Removal (RHR) 3.6.2.3 3.6.2.3 Suppression Pool Cooling One RHR suppression pool 3.6.2.3.A.1 3.6.2.3.A.1 Yes TSTF-505 changes are incorporated.
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| cooling subsystem inoperable.
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| In this Condition, the remaining RHR suppression pool cooling subsystem is adequate to perform the primary containment cooling function.
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| Suppression Pool Makeup 3.6.2.4 3.6.2.4 (SPMU) System One SPMU subsystem 3.6.2.4.C.1 3.6.2.4.C.1 Yes TSTF-505 changes are incorporated.
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| inoperable for reasons other than Condition A or B.
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| In this Condition, the redundant SPMU subsystem can perform the safety function.
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| GNRO-2023/00014 Page 14 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description TS GGNS TS RICT? Comments Drywell Air Lock 3.6.5.2 3.6.5.2 Drywell air lock inoperable for 3.6.5.2.C.3 3.6.5.2.C.2 Yes TSTF-505 changes are incorporated.
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| reasons other than Condition A GGNS Amendment 126 (ML021480466) moved the or B. drywell air lock leakage surveillance to TS 3.6.5.1, "Drywell," and deleted the specific overall leakage limit for the air lock leakage. The drywell air lock leakage rate limit does not reflect the ability of the drywell to perform its safety function. In this Condition, the ability of the drywell to perform its intended safety function is not dependent on the airlock seals meeting a specific leakage limit other than the total drywell leakage limit controlled by LCO 3.6.5.1, "Drywell."
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| Any potential loss of function resulting from drywell air lock leakage would be evaluated under LCO 3.6.5.1 and if inoperable, would require restoration within one hour. Therefore, there is no need to include a note that prohibits applying a RICT when a LOF has occurred.
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| Drywell Isolation Valves 3.6.5.3 3.6.5.3 One or more penetration flow 3.6.5.3.A.1 3.6.5.3.A.1 Yes TSTF-505 changes are incorporated.
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| paths with one drywell isolation valve inoperable. In this Condition, the remaining OPERABLE drywell isolation valve in the flow path is adequate to perform the isolation function.
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| GNRO-2023/00014 Page 15 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
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| Standby Service Water (SSW) 3.7.1 3.7.1 System and Ultimate Heat Sink (UHS)
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| One or more cooling towers with 3.7.1.A.1 3.7.1.A.1 Yes TSTF-505 changes are incorporated.
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| one cooling tower fan inoperable.
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| Condition A.1 applies when one UHS cooling tower has one inoperable fan. This leaves the remaining fan in the affected cooling tower and the other UHS cooling tower (with two fans) to perform the heat removal function.
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| One SSW subsystem inoperable. 3.7.1.C.1 3.7.1.D.1 Yes TSTF-505 changes are incorporated.
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| With the unit in this Condition, the remaining OPERABLE SSW subsystem is adequate to perform the heat removal function.
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| Main Turbine Bypass System 3.7.6 3.7.7 Requirements of the LCO not 3.7.6.A.1 3.7.7.A.1 No In this Condition, with the Main Turbine Bypass met or Main Turbine Bypass System inoperable, the assumptions of the design System inoperable. basis transient analysis may not be met. Therefore, this LCO is not eligible for the RICT program.
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| GNRO-2023/00014 Page 16 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
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| AC Sources - Operating 3.8.1 3.8.1 One required offsite circuit 3.8.1.A.3 3.8.1.A.2 Yes TSTF-505 changes are incorporated.
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| inoperable.
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| In this Condition, the remaining OPERABLE offsite circuit and DGs are adequate to supply electrical power to the onsite Class 1E distribution system.
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| One required DG inoperable. 3.8.1.B.4 3.8.1.B.4 Yes TSTF-505 changes are incorporated.
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| In this Condition, the remaining OPERABLE DGs and offsite circuits are adequate to supply electrical power to the onsite Class 1E distribution system.
| |
| Two required offsite circuits 3.8.1.C.2 3.8.1.C.2 Yes TSTF-505 changes are incorporated.
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| inoperable.
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| In this Condition, sufficient onsite AC sources are available to maintain the unit in a safe shutdown condition in the event of a DBA or transient.
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| One required offsite circuit 3.8.1.D.1 3.8.1.D.1 Yes TSTF-505 changes are incorporated.
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| inoperable. 3.8.1.D.2 3.8.1.D.2 In this Condition, the remaining OPERABLE offsite IAND circuit and DG are adequate to supply electrical One required DG inoperable. power to the onsite Class 1E distribution system.
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| One required automatic load 3.8.1.F.1 3.8.1.F.1 Yes In this Condition, the remaining OPERABLE sequencer inoperable. sequencer and power sources are adequate to supply electrical power to the onsite Class 1E distribution system.
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| GNRO-2023/00014 Page 17 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
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| DC Sources - Operating 3.8.4 3.8.4 One required battery charger on 3.8.4.A.3 ------------ No GGNS TS do not have the equivalent seven-day CT Division I or II inoperable. for restoring a battery charger to OPERABLE status. Instead, Action A.1 relies on verification of meeting battery cell parameters Category A limits.
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| One [or two] batter[y][ies on one 3.8.4.B.1 ------------ No GGNS TS do not have a Condition specific to division] inoperable. inoperable batteries, which would be addressed by Condition C.
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| Division I or II DC electrical 3.8.4.C.1 3.8.4.C.1 Yes TSTF-505 changes are incorporated.
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| power subsystem inoperable for reasons other than Condition A. In this Condition, the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition.
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| GNRO-2023/00014 Page 18 of 18 Table A4-1 Cross-Reference of TSTF-505 and GGNS Technical Specifications TSTF-505 Apply TS Description GGNS TS Comments Rev. 2 TS RICT?
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| Inverters - Operating 3.8.7 ------------
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| Division I or II inverter 3.8.7.A.1 ------------ No GGNS TS do not contain an LCO specific to inoperable. inverters.
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| Distribution Systems - 3.8.9 3.8.7 Operating One or more Division I or II AC 3.8.9.A.1 3.8.7.A.1 Yes TSTF-505 changes are incorporated.
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| electrical power distribution subsystems inoperable. In this Condition, the remaining AC electrical power distribution subsystems can support the minimum safety functions necessary to shut down the reactor and maintain it in a safe shutdown condition.
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| Condition E addresses a loss of function.
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| One or more Division I or II AC 3.8.9.B.1 ------------ No GGNS TS do not contain a Condition specific to AC vital bus distribution subsystems vital buses.
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| inoperable.
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| One or more Division I or II DC 3.8.9.C.1 3.8.7.B.1 Yes TSTF-505 changes are incorporated.
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| electrical power distribution subsystems inoperable. In this Condition, the remaining DC electrical power distribution subsystems can support the minimum safety functions necessary to shut down the reactor and maintain it in a safe shutdown condition.
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| Programs and Manuals 5.5 5.5 Risk Informed Completion Time 5.5.15 5.5.14 N/A TSTF-505 changes are incorporated.
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| Program
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| Attachment 5 GNRO-2023/00014 Evaluation of Instrumentation and Control Systems (44 pages follow)
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| GNRO-2023/00014 Page 1 of 44 Information Supporting Redundancy and Diversity This attachment provides justification for TS 3.3, Instrumentation, that defense-in-depth objectives are met and that at least one redundant or diverse means (through other automatic features or manual action) to accomplish necessary safety functions remains available during application of the RICT. This attachment identifies the components available to respond to identified accident conditions.
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| The following Instrumentation Technical Specifications (TS) sections are included in this TSTF-505 License Amendment Request (LAR) for Grand Gulf Nuclear Station (GGNS).
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| : 1. Reactor Protection System (RPS) Instrumentation - TS section 3.3.1.1
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| : 2. End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation - TS section 3.3.4.1
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| : 3. Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT)
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| Instrumentation - TS section 3.3.4.2
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| : 4. Emergency Core Cooling System (ECCS) Instrumentation - TS section 3.3.5.1
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| : 5. Reactor Core Isolation Cooling (RCIC) System Instrumentation - TS section 3.3.5.3
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| : 6. Primary Containment and Drywell Isolation Instrumentation - TS section 3.3.6.1
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| : 7. Residual Heat Removal (RHR) Containment Spray System Instrumentation- TS section 3.3.6.3
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| : 8. Relief and Low-Low Set (LLS) Instrumentation - LCO 3.3.6.5
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| : 9. Loss of Power (LOP) Instrumentation - TS section 3.3.8.1 GGNS TS 3.3 Instrumentation Limiting Conditions for Operation (LCOs) were developed to ensure that GGNS maintains necessary redundancy and diversity of systems, structures, and components (SSCs), and complies with the single failure design criterion as defined in IEEE 279-1971, and the diversity requirements as defined in Appendix A, General Design Criteria for Nuclear Power Plants to Part 50 of 10 CFR, GDC-22, Protection System Independence.
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| Included below is a description of the redundant and diverse means available to mitigate accidents that each identified instrumentation and control function defined in TS Section 3.3 is designed to prevent.
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| GNRO-2023/00014 Page 2 of 44 Each TS Instrumentation section in this attachment has a table that identifies the GGNS Updated Final Safety Analysis Report (UFSAR) transient/accident that credits the instrumentation and control function. For each instrumentation and control function, diverse instrumentation is identified to demonstrate that at least one diverse means is available to accomplish the associated safety function. The UFSAR, industry sources, and engineering judgement were used to identify transient/accidents that would challenge a response from the listed instrumentation. The list of potential accidents and redundant instrumentation are best estimates and are in no way exhaustive.
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| The following abbreviations are used within the Event column of the included tables:
| |
| * IMF-T: Incidents of moderate frequency. This event is referred to as an anticipated operational transient.
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| * II-T: Infrequent incidents. This event is referred to as an abnormal operational transient.
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| * DBA: Design basis accident/Limiting fault. This event is not expected to occur.
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| * ATWS: Anticipated transient without a scram. This event is not expected to occur.
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| GNRO-2023/00014 Page 3 of 44 A-1. Reactor Protection System (RPS) Instrumentation - TS Section 3.3.1.1 The RPS is comprised of two independent trip systems (A and B), with two logic channels in each trip system (i.e., logic channels A and C, B and D). The outputs of the logic channels in a trip system are combined in a one-out-of-two logic so either channel can trip the associated trip system. Trip of a system de-energizes the associated divisions scram solenoid on each control rod scram pilot valve. The tripping of both trip systems produces a reactor scram by de-energizing both scram solenoids on each control rod pilot valve which in turn opens the scram inlet and outlet valves aligning high pressure water to hydraulically cause rapid control rod insertion. This logic arrangement is referred to as one-out-of-two twice logic. Functions with a different logic are noted below.
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| The RPS design creates defense-in-depth from the redundancy of the channels for each trip system.
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| The diverse inputs causing a trip of the RPS are: (UFSAR Chapter 7.2, TS Table 3.3.1.1-1 and Bases B 3.3.1.1):
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| * 1.a. Intermediate Range Monitors - Neutron Flux - High o 4 channels per trip system: 8 channels per function
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| * 1.b. Intermediate Range Monitors - Inop o 4 channels per trip system: 8 channels per function
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| * 2.a. Average Power Range Monitors - Neutron Flux - High, Setdown o 4 channels, each providing input to each of the 4 2-Out-Of-4 Voter channels
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| * 2.b. Average Power Range Monitors - Fixed Neutron Flux - High o 4 channels, each providing input to each of the 4 2-Out-Of-4 Voter channels
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| * 2.c. Average Power Range Monitors - Inop o 4 channels, each providing input to each of the 4 2-Out-Of-4 Voter channels
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| * 2.d. Average Power Range Monitors - Flow Biased Simulated Thermal Power - High o 4 channels, each providing input to each of the 4 2-Out-Of-4 Voter channels
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| * 2.e. 2-Out-Of-4 Voter o 2 channels per trip system, 4 channels per function
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| * 2.f. OPRM Upscale o 4 channels, each providing input to each of the 4 2-Out-Of-4 Voter channels
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| * 3. Reactor Vessel Steam Dome Pressure - High o 2 channels per trip system: 4 channels per function
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| * 4. Reactor Vessel Water Level - Low, Level 3 o 2 channel per trip system: 4 channels per function
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| GNRO-2023/00014 Page 4 of 44
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| * 5. Reactor Vessel Water Level - High, Level 8 o 2 channel per trip system: 4 channels per function
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| * 6. Main Steam Isolation Valve - Closure o 1 channel for each trip-system per each MSIV: 16 channels per function
| |
| * 7. Drywell Pressure - High o 2 channels per trip system: 4 channels per function
| |
| * 8.a. SCRAM Discharge Volume Water Level - High - Transmitter/Trip Unit o 2 channels per trip system: 4 channels per function
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| * 8.b. SCRAM Discharge Volume Water Level - High - Float Switch o 2 channels per trip system: 4 channels per function
| |
| * 9. Turbine Stop Valve - Closure, Trip Oil Pressure - Low o 4 channels per trip system: 8 channels per function
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| * 10. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low o 2 channel per trip system: 4 channels per function
| |
| * 11. Reactor Mode Switch - Shutdown Position o 2 channels per trip system: 4 channels per function
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| * 12. Manual SCRAM o 2 channels per trip system: 4 channels per function In addition, GGNS has redundant and diverse methods of shutting down the reactor in the unlikely event that the RPS does not scram the reactor. The Alternate Rod Insertion (ARI) system provides backup capability to insert the control rods into the reactor and can be manually or automatically initiated.
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| GNRO-2023/00014 Page 5 of 44 Table 1 - RPS Instrumentation Redundancy/Diversity (TS Table 3.3.1.1-1)
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| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1
| |
| : 1. Intermediate Range Monitors (IRM) 1a. Neutron Flux - 15.1.6 Inadvertent 1) Automatic Initiation: IMF-T High Residual Heat - IRM Neutron Flux High Removal (RHR) - IRM Inop (8 channels, 4 per Shutdown - APRM Neutron Flux - High trip system, Cooling (Setdown) 1-out-of-4 taken Operation 2) Manual scram twice logic) 15.4.1 Rod Withdrawal 1) Automatic Initiation: II-T Error - Low - IRM Neutron Flux High Power - IRM Inop
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| - Rod Control Information System (RCIS) Banked Position Withdraw Sequence (BPWS)
| |
| - APRM Fixed Neutron Flux -
| |
| High
| |
| - APRM Neutron Flux - High (Setdown)
| |
| : 2) Manual scram 1b. Inop None None 1) Manual scram N/A (8 channels, 4 per The IRM Inop logic is designed trip system, to protect against multiple IRMs arranged in a inoperable. It provides an 1-out-of-4 taken automatic trip signal if multiple twice logic) IRMs are inoperable.
| |
| : 2. Average Power Range Monitors (APRM) 2a. Neutron Flux - 15.1.6 Inadvertent RHR 1) Automatic Initiation IMF-T High Setdown Shutdown - IRM Neutron Flux - High Cooling - APRM Neutron Flux - High (4 channels and 4 Operation (Setdown)
| |
| Voter channels. - APRM Inop Each APRM 2) Manual scram channel provides 15.4.9 Control Rod 1) Automatic Initiation DBA input to each of Drop Accident - IRM Neutron Flux - High the four voter (CRDA) - APRM Neutron Flux - High channels.) (Setdown)
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| - APRM Inop
| |
| - RCIS BPWS (Rod control and information system banked position withdrawal sequencing)
| |
| : 2) Manual scram
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| GNRO-2023/00014 Page 6 of 44 Table 1 - RPS Instrumentation Redundancy/Diversity (TS Table 3.3.1.1-1)
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| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1 2b. Fixed Neutron 15.2.1 Pressure 1) Automatic Initiation: IMF-T Flux - High Controller - APRM Flow Biased Sim.
| |
| Failure - Closed Therm. Power - High (4 channels and 4 - APRM Neutron Flux - High Voter channels. (Setdown)
| |
| Each APRM - APRM Fixed Neutron Flux -
| |
| channel provides High input to each of - APRM Inop the four voter - Reactor Vessel Steam Dome channels.) Pressure - High
| |
| : 2) Manual scram 15.2.4 Main Steam 1) Automatic Initiation IMF-T Isolation Valve - APRM Flow Biased Sim.
| |
| (MSIV) Closures Therm. Power - High
| |
| - APRM Fixed Neutron Flux -
| |
| High
| |
| - APRM Inop
| |
| - MSIV -Closure
| |
| - Reactor Steam Dome Pressure - High
| |
| - Reactor Vessel Water Level Low Level 3
| |
| : 2) Manual scram 15.4.9 CRDA 1) Automatic Initiation: DBA
| |
| - APRM Fixed Neutron Flux -
| |
| High
| |
| - APRM Inop
| |
| - IRM Neutron Flux - High
| |
| - RCIS BPWS
| |
| : 2) Manual scram 2c. Inop None None 1) Manual scram N/A (4 channels and 4 The APRM Inop logic is Voter channels. designed to protect against Each APRM multiple APRMs inoperable. It channel provides provides an automatic trip input to each of signal if multiple APRMs are the four voter inoperable.
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| channels.)
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| GNRO-2023/00014 Page 7 of 44 Table 1 - RPS Instrumentation Redundancy/Diversity (TS Table 3.3.1.1-1)
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| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1 2d. Flow Biased 15.1.1 Loss of 1) Automatic Initiation IMF-T Simulated Feedwater - APRM Flow Biased Sim.
| |
| Thermal Power - Heater, Therm. Power - High High automatic flow - APRM Fixed Neutron Flux -
| |
| control High (4 channels and 4 - APRM Inop Voter channels. - Turbine Stop Valve - Closure Each APRM 2) Manual scram channel provides input to each of the four voter channels.)
| |
| 2e. 2-Out-Of-4 None None Provides the interface between N/A Voter the APRM functions, including the OPRM Upscale (Four channels) function, and the final RPS trip system logic.
| |
| 2f. OPRM Upscale 15.B.3 General Design Complies with GDC 10 and 12. N/A Criteria (GDC) Provides protection from fuel (4 channels and 4 10 and 12 (10 minimum critical power ratio Voter channels. CFR 50, (MCPR) safety limit (SL) due Each APRM Appendix A) to anticipated thermal-channel provides hydraulic power oscillations.
| |
| input to each of the four voter channels.)
| |
| : 3. Reactor Vessel 15.2.1 Pressure 1) Automatic Initiation IMF-T Steam Dome Controller - APRM Flow Biased Sim.
| |
| Pressure - High Failure - Closed Therm. Power - High 15.2.2 Generator Load - APRM Neutron Flux - High IMF-T (4 channels, 2 per Rejection (Setdown) trip system, 15.2.3 Turbine Trip - APRM Fixed Neutron Flux - IMF-T arranged in a 1- High 15.2.4 MSIV Closures IMF-T out-of-2 taken - Reactor Vessel Steam Dome twice logic) 15.2.5 Loss of Pressure - High IMF-T Condenser - Reactor Vessel Water Level -
| |
| Vacuum Low, Level 3 15.2.6 Loss of AC 2) Manual Initiation IMF-T Power (Alternating current)
| |
| | |
| GNRO-2023/00014 Page 8 of 44 Table 1 - RPS Instrumentation Redundancy/Diversity (TS Table 3.3.1.1-1)
| |
| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1
| |
| : 4. Reactor Vessel 15.2.7 Loss of 1) Automatic Initiation: IMF-T Water Level - Feedwater Flow - Reactor Vessel Water Level -
| |
| Low, Level 3 Low Level 3
| |
| - Main Steam Isolation Valve -
| |
| (4 channels, 2 per Closure trip system, 2) Manual scram arranged in a 1- 15.6.5 Loss of Coolant 1) Automatic Initiation: DBA out-of-2 taken Accidents - Reactor Vessel Water Level -
| |
| twice logic) Low Level 3
| |
| - Primary Containment Pressure - High
| |
| : 2) Manual scram 15.2.8/ Feedwater Line 1) Automatic Initiation: DBA 15.6.6 Break - Outside - Reactor Vessel Water Level -
| |
| Containment Low Level 3
| |
| - Main Steam Isolation Valve -
| |
| Closure
| |
| : 2) Manual scram
| |
| : 5. Reactor Vessel 15.1.2 Feedwater 1) Automatic Initiation: IMF-T Water Level - Controller - Reactor Vessel Water Level -
| |
| High, Level 8 Failure - High, Level 8 Maximum - APRM Flow Biased Sim.
| |
| (4 channels, 2 per Demand Therm. Power - High trip system, - APRM Fixed Neutron Flux -
| |
| arranged in a 1- High out-of-2 taken - Reactor Vessel Steam Dome twice logic) Pressure - High
| |
| - Turbine Stop Valve - Closure
| |
| - Turbine Control Valve Fast Closure, Trip Oil Pressure -
| |
| Low
| |
| : 2) Manual scram 15.3.1 Recirculation 1) Automatic Initiation: IMF-T Pump Trip - Reactor Vessel Water Level -
| |
| High, Level 8
| |
| - Main Steam Isolation Valve -
| |
| Closure
| |
| - Turbine Stop Valve - Closure
| |
| : 2) Manual scram
| |
| | |
| GNRO-2023/00014 Page 9 of 44 Table 1 - RPS Instrumentation Redundancy/Diversity (TS Table 3.3.1.1-1)
| |
| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1
| |
| : 6. Main Steam 15.2.4 MSIV Closures 1) Automatic Initiation: IMF-T Isolation Valve - - Main Steam Isolation Valve -
| |
| Closure Closure
| |
| - APRM Flow Biased Sim.
| |
| (16 channels, 1 Therm. Power - High channel for each - Reactor Vessel Steam Dome trip-system per Pressure - High each MSIV, 2) Manual scram arranged in a 2- 15.6.4 Steam System 1) Automatic Initiation DBA out-of-8 taken Piping Break - Main Steam Isolation Valve -
| |
| twice on a Outside Closure minimum of three Containment - Reactor Vessel Water Level -
| |
| MS lines logic) Low Level 3
| |
| : 2) Manual scram
| |
| : 7. Drywell 15.6.5 Loss of Coolant 1) Automatic Initiation: DBA Pressure - High Accidents - Reactor Vessel Water Level -
| |
| Low, Level 3 (4 channels, 2 per - Drywell Pressure - High trip system, 2) Manual scram arranged in a 1-out-of-2 taken twice logic)
| |
| : 8. Scram Discharge Volume Water Level (SDVWL) - High 8a. None None 1) Automatic Initiation N/A Transmitter/Trip - SDVWL Transmitter/Level Unit Indicating Switch
| |
| - SDVWL Float Switch (4 channels, 2 per 2) Manual SCRAM trip system, arranged in a 1-out-of-2 taken twice logic) 8b. Float Switch None None 1) Automatic Initiation N/A
| |
| - SDVWL Transmitter/Level (4 channels, 2 per indicating Switch trip system, - SDVWL Float Switch arranged in a 1- 2) Manual SCRAM out-of-2 taken twice logic)
| |
| | |
| GNRO-2023/00014 Page 10 of 44 Table 1 - RPS Instrumentation Redundancy/Diversity (TS Table 3.3.1.1-1)
| |
| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1
| |
| : 9. Turbine Stop 15.1.2 Feedwater 1) Automatic Initiation: IMF-T Valve Closure Trip Controller - Main Steam Isolation Valve -
| |
| Oil Pressure - Failure - Closure Low Maximum - APRM Flow Biased Sim.
| |
| Demand Therm. Power - High (8 channels, 4 per 15.2.3 Turbine Trip - APRM Fixed Neutron Flux - IMF-T trip system, 15.2.5 Loss of High IMF-T arranged in a 2- Condenser - Turbine Stop Valve - Closure out-of-4 taken Vacuum - Reactor Vessel Steam Dome twice logic) Pressure - High
| |
| : 2) Manual scram 15.3.1 Recirculation 1) Automatic Initiation: IMF-T Pump Trip - Main Steam Isolation Valve -
| |
| Closure
| |
| - Turbine Stop Valve - Closure
| |
| : 2) Manual scram 15.3.2 Recirculation 1) Automatic Initiation: IMF-T Flow Control - Main Steam Isolation Valve -
| |
| Failure - Closure Decreasing Flow - Reactor Vessel Steam Dome 15.3.3 Recirculation Pressure - High DBA Pump Seizure - Turbine Stop Valve - Closure 15.3.4 Recirculation 2) Manual scram DBA Pump Shaft Break
| |
| | |
| GNRO-2023/00014 Page 11 of 44 Table 1 - RPS Instrumentation Redundancy/Diversity (TS Table 3.3.1.1-1)
| |
| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1
| |
| : 10. Turbine 15.1.2 Feedwater 1) Automatic Initiation: IMF-T Control Valve Fast Controller - APRM Flow Biased Sim.
| |
| Closure, Trip Oil Failure - Therm. Power - High Pressure - Low Maximum - APRM Fixed Neutron Flux -
| |
| Demand High (4 channels, 2 per 15.2.2 Generator Load - Reactor Vessel Steam Dome IMF-T trip system, Rejection Pressure - High arranged in a 1- - TSV Closure out-of-2 taken - TCV Fast Closure twice logic) 2) Manual scram 15.2.6 Loss of AC 1) Automatic Initiation IMF-T Power - APRM Flow Biased Sim.
| |
| Therm. Power - High
| |
| - APRM Neutron Flux - High (Setdown)
| |
| - APRM Fixed Neutron Flux -
| |
| High
| |
| - Reactor Vessel Steam Dome Pressure - High
| |
| - Reactor Vessel Water Level -
| |
| Low, Level 3
| |
| - TCV Fast Closure
| |
| : 2) Manual Initiation
| |
| : 11. Reactor Mode None None 1) Automatic Initiation N/A Switch - - See Functions 1 through 9 Shutdown Position 2) Manual scram (4 channels, 2 per trip system, arranged in a 1-out-of-2 taken twice logic)
| |
| : 12. Manual Scram None None 1) Automatic Initiation N/A
| |
| - See Functions 1 through 9 (4 channels, 2 per 2) Reactor Mode Switch -
| |
| trip system, Shutdown Position arranged in a 1-out-of-2 taken twice logic)
| |
| Note 1: Instrumentation in italics is the affected function crediting redundant channels.
| |
| | |
| GNRO-2023/00014 Page 12 of 44 A-2. End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation - TS Section 3.3.4.1 The EOC-RPT instrumentation initiates a recirculation pump trip (RPT) to reduce the peak reactor pressure and power resulting from turbine trip or generator load rejection transients to provide additional margin to core thermal minimum critical power ratio (MCPR) safety limits. The need for the additional negative reactivity in excess of that normally inserted on a scram reflects end of cycle reactivity considerations. Reactor scram functions are not considered as diverse redundant functions.
| |
| The EOC-RPT has two identical trip systems, either of which can actuate an RPT. Each EOC-RPT system is a 2-out-of-2 initiation logic for the Trip Function. If either trip system actuates, both recirculation pumps will trip.
| |
| The EOC-RPT design creates defense-in-depth from the redundancy of trip systems for the Trip Function.
| |
| Diverse EOC-RPT systems:
| |
| * a.1. Turbine Stop Valve (TSV) Closure, Trip Oil Pressure - Low:
| |
| o 4 instrument channels arranged in a 2-out-of-2 logic in either trip system.
| |
| * a.2. Turbine Control Valve (TCV) Fast closure, Trip Oil Pressure - Low:
| |
| o 4 instrument channels arranged in a 2-out-of-2 logic in either trip system.
| |
| Table 2 - EOC-RPT Instrumentation Redundancy/Diversity (LCO 3.3.4.1)
| |
| UFSAR Redundant/Diverse Function Transient Accident Event Section Instrumentation1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation a.1. Turbine Stop 15.1.2 Feedwater Controller 1) Automatic Initiation: IMF-T Valve (TSV) - Failure - Maximum - TSV Closure Closure, Trip Oil Demand - TCV Fast closure, Trip Pressure - Low 15.2.2 Generator Load Reject Oil Pressure - Low IMF-T 15.2.3 Turbine Trip 2) Manual RPT IMF-T (4 channels, 2 per 15.2.5 Loss of condenser IMF-T trip system, Vacuum arranged in a 2-out- 15.2.6 Loss of AC Power IMF-T of-2 logic in either trip system)
| |
| | |
| GNRO-2023/00014 Page 13 of 44 Table 2 - EOC-RPT Instrumentation Redundancy/Diversity (LCO 3.3.4.1)
| |
| UFSAR Redundant/Diverse Function Transient Accident Event Section Instrumentation1 a.2. Turbine 15.1.2 Feedwater Controller 1) Automatic Initiation: IMF-T Control Valve Failure - Maximum - TSV Closure (TCV) Fast closure, Demand - TCV Fast closure, Trip Trip Oil Pressure - 15.2.2 Generator Load Reject Oil Pressure - Low IMF-T Low 15.2.3 Turbine Trip 2) Manual RPT IMF-T 15.2.5 Loss of condenser IMF-T (4 channels, 2 per Vacuum trip system, 15.2.6 Loss of AC Power IMF-T arranged in a 2-out-of-2 logic in either trip system)
| |
| Note 1: Instrumentation in italics is the affected function crediting redundant channels
| |
| | |
| GNRO-2023/00014 Page 14 of 44 A-3. Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT)
| |
| Instrumentation - TS Section 3.3.4.2 The ATWS-RPT consists of two independent trip systems, with two channels of the Reactor Vessel Pressure-High and two channels of Reactor Vessel Water Level-Low Low, Level 2, in each trip system. Each ATWS-RPT trip system is a 2-out-of-2 logic for either function. Tripping either trip system will trip all recirculation pump breakers. There is one fast speed motor breaker and one low frequency motor generator motor breaker provided for each of the two recirculation pumps for a total of four breakers. The output of each trip system is provided to all four breakers.
| |
| The ATWS-RPT design creates defense-in-depth from the redundancy of trip systems for the Trip Function.
| |
| Diverse ATWS-RPT trip systems:
| |
| * a. Reactor Vessel Water Level - Low Low, Level 2:
| |
| o 4 instrument channels, 2 channels per trip system.
| |
| * b. Reactor Vessel Pressure - High:
| |
| o 4 instrument channels, 2 channels per trip system.
| |
| | |
| GNRO-2023/00014 Page 15 of 44 Table 3 - ATWS-RPT Instrumentation Redundancy/Diversity (LCO 3.3.4.2)
| |
| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation 1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT)
| |
| Instrumentation
| |
| : a. Reactor Vessel 15.8 Anticipated 1) Automatic Initiation: ATWS Water Level - Low transients - Reactor Vessel Water Level Low, Level 2 Without SCRAM - Low Low, Level 2
| |
| - Reactor Vessel Steam (4 channels, 2 per trip Dome Pressure - High system, arranged in a 2) Manual RPT 2-out-of 2 in either trip system logic)
| |
| : b. Reactor Vessel 15.8 Anticipated 1) Automatic Initiation: ATWS Steam Dome transients - Reactor Vessel Water Level Pressure - High Without SCRAM - Low Low, Level 2
| |
| - Reactor Vessel Steam (4 channels, 2 per trip Dome Pressure - High system, arranged in a 2) Manual RPT 2-out-of 2 in either trip system logic)
| |
| Note 1: Instrumentation in italics is the affected function crediting redundant channels.
| |
| | |
| GNRO-2023/00014 Page 16 of 44 A-4. Emergency Core Cooling System (ECCS) Instrumentation - TS Section 3.3.5.1 The purpose of the ECCS instrumentation is to initiate appropriate responses from the systems to ensure that fuel is adequately cooled in the event of a design basis accident or transient. The ECCS instrumentation actuates low pressure core spray (LPCS), low pressure coolant injection (LPCI), high pressure core spray (HPCS), Automatic Depressurization System (ADS), and the diesel generators (DGs).
| |
| The ECCS Instrumentation design creates defense-in-depth from the redundancy of the channels for each trip system performing the Trip Function (ECCS Actuation).
| |
| Diverse ECCS instrumentation inputs (TS Table 3.3.5.1-1 and TS Basis B 3.3.5.1):
| |
| * 1. Low Pressure Coolant Injection-A (LPCI) and Low Pressure Core Spray (LPCS) Subsystems o 1.a. Reactor Vessel Water Level - Low Low Low, Level 1 1.b. Drywell Pressure - High:
| |
| * Reactor vessel water level (trip level 1) is monitored by two redundant differential pressure switches. Drywell pressure is monitored by two redundant pressure switches. The switches are connected in a 1-out-of-2 taken twice logic.
| |
| o 1.c. LPCI Pump A Start - Time Delay Relay:
| |
| * There is one time delay relay arranged in a single channel logic for the pump.
| |
| o 1.d. Reactor Vessel Pressure - Low (Injection Permissive):
| |
| * There are two pressure transmitters that sense the reactor dome pressure. Each transmitter provides input to two pressure switch trip units (four channels total) for both Division A low pressure ECCS injection valves, LPCS and LPCIA. These four channels are arranged in a one-out-of-two taken twice logic for both injection valves.
| |
| o 1.e. LPCS Pump Discharge Flow - Low (Bypass):
| |
| * One flow indicating switch arranged in a single channel logic.
| |
| o 1.f. LPCI Pump A Discharge Flow - Low (Bypass):
| |
| * One flow indicating switch arranged in a single channel logic.
| |
| o 1.g. Manual Initiation:
| |
| * For each low pressure ECCS actuation system, there is one push button arranged in a single channel logic.
| |
| * 2. LPCI B and LPCI C Subsystems o 2.a. Reactor Vessel Water Level - Low Low Low, Level 1 2.b. Drywell Pressure - High:
| |
| | |
| GNRO-2023/00014 Page 17 of 44 Reactor vessel water level (trip level 1) is monitored by two redundant differential pressure switches. Drywell pressure is monitored by two redundant pressure switches. The switches are connected in a one-out-of-two taken twice logic.
| |
| o 2.c. LPCI Pump B Start - Time Delay Relay:
| |
| There is one time delay relay arranged in a single channel logic for each pump.
| |
| o 2.d. Reactor Vessel Pressure - Low (Injection Permissive):
| |
| There are two pressure transmitters that sense the reactor dome pressure. Each transmitter provides input to two pressure switch trip units (four channels total) for both Division B low pressure ECCS injection valves, LPCIB and LPCIC. These four channels are arranged in a one-out-of-two taken twice logic for both injection valves.
| |
| o 2.e. LPCI Pump B and LPCI Pump C Discharge Flow - Low (Bypass):
| |
| One flow indicating switch per pump, arranged in a single channel logic.
| |
| o 2.f. Manual Initiation:
| |
| For each low pressure ECCS actuation system, there is one push button arranged in a single channel logic.
| |
| * 3. High Pressure Core Spray (HPCS) System o 3.a. Reactor Vessel Water Level - Low Low, Level 2:
| |
| 4 instrument channels arranged in a one-out-of-two taken twice logic.
| |
| o 3.b. Drywell Pressure - High:
| |
| 4 instrument channels arranged in a one-out-of-two taken twice logic.
| |
| o 3.c. Reactor Vessel Water Level - High, Level 8:
| |
| 2 instrument channels arranged in a two-out-of-two logic.
| |
| o 3.d. Condensate Storage Tank Level - Low:
| |
| 2 instrument channels arranged in a one-out-of-two logic.
| |
| o 3.e. Suppression Pool Water Level - High:
| |
| 2 instrument channels arranged in a one-out-of-two logic.
| |
| o 3.f. HPCS Pump Discharge Pressure High (Bypass):
| |
| 1 instrument channel arranged in a single channel logic.
| |
| o 3.g. HPCS System Flow Rate - Low (Bypass):
| |
| 1 instrument channel arranged in a single channel logic.
| |
| o 3.h. Manual Initiation:
| |
| One push button arranged in a single channel logic.
| |
| * 4&5. Automatic Depressurization System (ADS) Trip System A and B o 4.a. Reactor Vessel Water Level - Low Low Low, Level 1 5.a. Reactor Vessel Water Level - Low Low Low, Level 1:
| |
| | |
| GNRO-2023/00014 Page 18 of 44 2 independent trip systems A and B. 2 channels per trip system arranged in a 2-out-of-2 logic for each trip system.
| |
| o 4.b. Drywell Pressure - High 5.b. Drywell Pressure - High:
| |
| 2 independent trip systems A and B. 2 channels per trip system arranged in a 2-out-of-2 logic for each trip system.
| |
| o 4.c. ADS Initiation Timer 5.c. ADS Initiation Timer:
| |
| 2 channels with one in each trip system A and B arranged in a single channel logic.
| |
| o 4.d. Reactor Vessel Water Level - Low Level 3 (Confirmatory) 5.d. Reactor Vessel Water Level - Low Level 3 (Confirmatory):
| |
| 2 channels with one in each trip system A and B arranged in a single channel logic.
| |
| o 4.e. LPCS Pump Discharge Pressure - High 4.f. LPCI Pump A Discharge Pressure - High 5.e. LPCI Pumps B & C Discharge Pressure - High:
| |
| 2 LPCS and 2 LPCI A channels input to ADS trip system A. 2 LPCI B and 2 LPC C channels input to ADS trip system B. In order to generate an ADS permissive, one trip system, with only one pump (both channels for the pump) must indicate the high discharge pressure condition.
| |
| o 4.g. ADS Bypass Timer (High Drywell Pressure) 5.f. ADS Bypass Timer (High Drywell Pressure):
| |
| 4 channels with two in each trip system A and B arranged in a 2-out-of-2 logic.
| |
| o 4.h. Manual Initiation 5.g. Manual Initiation 2 independent trip systems A and B. 2 push buttons per ADS trip system in a 2-out-of-2 logic.
| |
| | |
| GNRO-2023/00014 Page 19 of 44 Table 4 - ECCS Instrumentation Redundancy/Diversity (TS Table 3.3.5.1-1)
| |
| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1
| |
| : 1. Low Pressure Coolant Injection-A (LPCI) and Low Pressure Core Spray (LPCS)
| |
| Subsystems 1.a. Reactor Vessel 15.6.5 Loss of Coolant 1) Automatic Initiation: DBA Water Level - Low Accidents - LPCI A & LPCS Initiation Low Low, Level 1 Drywell Pressure - High
| |
| - LPCI B & C Initiation Reactor (2 channels, 1 per trip Vessel Water Level - Low system, arranged in a Low Low, Level 1 1-out-of-2 taken twice - LPCI B & C Initiation Drywell logic in combination Pressure - High with function 1.b) - HPCS Initiation Drywell Pressure - High
| |
| - HPCS Initiation Reactor Vessel Water Level - Low Low, Level 2
| |
| : 2) Manual Initiation 1.b. Drywell Pressure 15.6.5 Loss of Coolant 1) Automatic Initiation: DBA
| |
| - High Accidents - LPCI A & LPCS Reactor Vessel Water Level - Low (2 channels, 1 per trip Low Low, Level 1 system, arranged in a - LPCI B & C Initiation Reactor 1-out-of-2 taken twice Vessel Water Level - Low logic in combination Low Low, Level 1 with function 1.a.) - LPCI B & C Initiation Drywell Pressure - High
| |
| - HPCS Initiation Drywell Pressure - High
| |
| - HPCS Initiation Reactor Vessel Water Level - Low Low, Level 2
| |
| : 2) Manual Initiation 1.c. LPCI Pump A Implicitly assumed for LPCS Pump Initiation NA Start - Time Delay accidents that initiate LPCI LPCI Pump B Initiation Relay Pump A LPCI Pump C Initiation (single channel logic) 1.d. Reactor Vessel Implicitly assumed for LPCI Pump B Initiation N/A Pressure - Low accidents that initiate LPCS LPCI Pump C Initiation (Injection Permissive) Pump or LPCI Pump A (4 channels, 2 per trip sub-system, arranged in a 1-out-of-2 taken twice logic)
| |
| | |
| GNRO-2023/00014 Page 20 of 44 Table 4 - ECCS Instrumentation Redundancy/Diversity (TS Table 3.3.5.1-1)
| |
| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1 1.e. LPCS Pump Implicitly assumed for LPCI Pump A initiation N/A Discharge Flow - Low accidents that initiate LPCS LPCI Pump B initiation (Bypass) Pump LPCI Pump C initiation (single channel logic) 1.f. LPCI Pump A Implicitly assumed for LPCS Pump initiation N/A Discharge Flow - Low accidents that initiate LPCI LPCI Pump B initiation (Bypass) Pump A LPCI Pump C initiation (single channel logic) 1.g. Manual Initiation None None 1) Automatic Initiation: N/A
| |
| - See Functions 1.a and 1.b (single channel logic) and 2.a and 2.b
| |
| : 2) Manual Initiation
| |
| - Function 2.f to initiate LPCI B and LPCI C.
| |
| : 2. LPCI B and LPCI C Subsystems 2.a. Reactor Vessel 15.6.5 Loss of Coolant 1) Automatic Initiation: DBA Water Level - Low Accidents - LPCI B & LPCI C Initiation Low Low, Level 1 Drywell Pressure - High
| |
| - LPCI A & LPCS Initiation (2 channels, 1 per trip Reactor Vessel Water Level system, arranged in a - Low Low Low, Level 1 1-out-of-2 taken twice - LPCI A & LPCS Initiation logic in combination Drywell Pressure - High with function 2.b.) - HPCS Initiation Reactor Vessel Water Level - Low Low, Level 2
| |
| - HPCS Initiation Drywell Pressure - High
| |
| : 2) Manual Initiation
| |
| | |
| GNRO-2023/00014 Page 21 of 44 Table 4 - ECCS Instrumentation Redundancy/Diversity (TS Table 3.3.5.1-1)
| |
| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1 2.b. Drywell Pressure 15.6.5 Loss of Coolant 1) Automatic Initiation: DBA
| |
| - High Accidents - LPCI B & C Initiation Reactor Vessel Water Level (2 channels, 1 per trip - Low Low Low, Level 1 system, arranged in a - LPCI A & LPCS Initiation 1-out-of-2 taken twice Reactor Vessel Water Level logic in combination - Low Low Low, Level 1 with function 2.a.) - LPCI A & LPCS Initiation Drywell Pressure - High
| |
| - HPCS Initiation Reactor Vessel Water Level - Low Low, Level 2
| |
| - HPCS Initiation Drywell Pressure - High
| |
| : 2) Manual Initiation 2.c. LPCI Pump B Implicitly assumed for LPCS Pump initiation N/A Start - Time Delay accidents that initiate LPCI LPCI Pump A initiation Relay Pump B LPCI Pump C initiation (single channel logic) 2.d. Reactor Vessel Implicitly assumed for LPCS Pump initiation N/A Pressure - Low accidents that initiate LPCI LPCI Pump A initiation (Injection Permissive) Pump B or LPCI Pump C (4 channels, 2 per trip sub-system, arranged in a 1-out-of-2 taken twice logic) 2.e. LPCI Pump B and Implicitly assumed for LPCS Pump initiation N/A LPCI Pump C accidents that initiate LPCI LPCI Pump A initiation Discharge Flow - Low Pump B or LPCI Pump C LPCI Pump B or C initiation (Bypass) (operable train)
| |
| (single channel logic) 2.f. Manual Initiation None None 1) Automatic Initiation: N/A
| |
| - See Functions 1.a and 1.b (single channel logic) and 2.a and 2.b
| |
| : 2) Manual Initiation
| |
| - Function 1.g to initiate LPCS and LPCI A.
| |
| | |
| GNRO-2023/00014 Page 22 of 44 Table 4 - ECCS Instrumentation Redundancy/Diversity (TS Table 3.3.5.1-1)
| |
| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1
| |
| : 3. High Pressure Core Spray (HPCS) System 3.a. Reactor Vessel 15.1.2 Feedwater 1) Automatic Initiation: IMF-T Water Level - Low Controller Failure - - RCIC Initiation Low Low Low, Level 2 Maximum Demand Level 2 15.1.3 Pressure Controller - LPCI A & LPCS Initiation IMF-T (4 channels, 2 per trip Failure - Open Reactor Vessel Water Level system, arranged in a Low Low Low, Level 1 1-out-of-2 taken twice 15.2.1 Pressure Controller - LPCI B & C Initiation IMF-T logic) Failure - Closed Reactor Vessel Water Level 15.2.3 Turbine Trip - Low Low Low, Level 1 IMF-T 15.2.4 MSIV Closure 2) Manual Initiation IMF-T 15.2.7 Loss of Feedwater IMF-T Flow 15.3.4 Recirculation Pump DBA Shaft Break 15.6.4 Steam system DBA Piping Break Outside Containment 15.2.8/ Feedwater Line DBA 15.6.6 Break - Outside Containment 15.8 Loss of Feedwater ATWS ATWS 15.6.5 Loss of Coolant 1) Automatic Initiation: DBA Accidents - HPCS Initiation Drywell Pressure - High
| |
| - RCIC Initiation Low Low, Level 2
| |
| - LPCI A & LPCS Initiation Reactor Vessel Water Level
| |
| - Low Low Low, Level 1
| |
| - LPCI A & LPCS Initiation Drywell Pressure - High
| |
| - LPCI B & C Initiation Reactor Vessel Water Level
| |
| - Low Low Low, Level 1
| |
| - LPCI B & C Initiation Drywell Pressure - High
| |
| : 2) Manual Initiation
| |
| | |
| GNRO-2023/00014 Page 23 of 44 Table 4 - ECCS Instrumentation Redundancy/Diversity (TS Table 3.3.5.1-1)
| |
| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1 3.b. Drywell Pressure 15.6.5 Loss of Coolant 1) Automatic Initiation: DBA
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| - High Accidents - HPCS initiation Reactor Vessel Water Level - Low (4 channels, 2 per trip Low, Level 2 system, arranged in a - RCIC Initiation Low Low, 1-out-of-2 taken twice Level 2 logic) - LPCI A & LPCS Initiation Reactor Vessel Water Level
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| - Low Low Low, Level 1
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| - LPCI A & LPCS Initiation Drywell Pressure - High
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| - LPCI B & C Initiation Reactor Vessel Water Level
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| - Low Low Low, Level 1
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| - LPCI B & C Initiation Drywell Pressure - High
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| : 2) Manual Initiation 3.c. Reactor Vessel None None 1) Manual Trip None Water Level - High, Level 8 (2 channels arranged in a 2-out-of-2 logic) 3.d. Condensate Implicitly assumed for 1) Automatic Initiation: N/A Storage Tank Level - accidents that initiate HPCS - Suppression Pool Water Low Level - High
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| : 2) Manual Alignment (2 channels, arranged in a 1-out-of-2 logic) 3.e. Suppression Pool Implicitly assumed for 1) Automatic Initiation: N/A Water Level - High accidents that initiate HPCS - Condensate Storage Tank Level - Low (2 channels, arranged 2) Manual Alignment in a 1-out-of-2 logic) 3.f. HPCS Pump Implicitly assumed for 1) Manual Alignment N/A Discharge Pressure - accidents that initiate HPCS High (Bypass)
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| (single channel logic) 3.g. HPCS System Implicitly assumed for 1) Manual Alignment N/A Flow Rate - Low accidents that initiate HPCS (Bypass)
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| (single channel logic)
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| GNRO-2023/00014 Page 24 of 44 Table 4 - ECCS Instrumentation Redundancy/Diversity (TS Table 3.3.5.1-1)
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| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1 3.h. Manual Initiation None None 1) Automatic Initiation: N/A
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| - See Functions 3.a and 3.b (single channel logic)
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| : 4. Automatic Depressurization System (ADS) Trip System A 4.a. Reactor Vessel 15.6.5 Loss of Coolant 1) Automatic Initiation DBA Water Level - Low Accidents - ADS A Initiation Drywell Low Low, Level 1 Pressure - High
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| - ADS B Initiation Drywell (2 channels, arranged Pressure - High in a 2-out-of-2 logic for - ADS B Initiation Reactor A trip system) Vessel Water Level - Low Low Low, Level 1
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| : 2) Manual ADS Initiation 4.b. Drywell Pressure 15.6.5 Loss of Coolant 1) Automatic Initiation DBA
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| - High Accidents - ADS A Initiation Reactor Vessel Water Level - Low (2 channels arranged Low Low, Level 1 in a 2-out-of-2 logic for - ADS B Initiation Drywell A trip system) Pressure - High
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| - ADS B Initiation Reactor Vessel Water Level - Low Low Low, Level 1
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| : 2) Manual ADS Initiation 4.c. ADS Initiation 15.6.5 Loss of Coolant 1) Automatic Initiation DBA Timer Accidents - ADS B Trip System
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| : 2) Manual ADS Initiation (single channel logic) 4.d. Reactor Vessel 15.6.5 Loss of Coolant 1) Automatic Initiation DBA Water Level - Low Accidents - ADS B Trip System Level 3 (Confirmatory) 2) Manual ADS Initiation (single channel logic) 4.e. LPCS Pump 15.6.5 Loss of Coolant 1) Automatic Initiation DBA Discharge Pressure - Accident - LPCI Pump A Discharge High Pressure - High
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| - ADS B Trip System (2-out-of-2 channels 2) Manual ADS Initiation on one pump)
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| GNRO-2023/00014 Page 25 of 44 Table 4 - ECCS Instrumentation Redundancy/Diversity (TS Table 3.3.5.1-1)
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| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1 4.f. LPCI Pump A 15.6.5 Loss of Coolant 1) Automatic Initiation DBA Discharge Pressure - Accidents - LPCS Pump Discharge High Pressure - High
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| - ADS B Trip System (2-out-of-2 channels 2) Manual ADS Initiation on one pump) 4.g. ADS Bypass 15.6.5 Loss of Coolant 1) Automatic Initiation DBA Timer (High Drywell Accidents - ADS B Trip System Pressure) 2) Manual ADS Initiation (2-out-of-2 channels) 4.h. Manual Initiation None None 1) Automatic Initiation N/A
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| - ADS A Initiation Reactor (2 push buttons per Vessel Water Level - Low ADS A trip system in a Low Low, Level 1 2-out-of-2 logic) - ADS B Initiation Reactor Vessel Water Level - Low Low Low, Level 1
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| : 5. ADS Trip System B 5.a. Reactor Vessel 15.6.5 Loss of Coolant 1) Automatic Initiation DBA Water Level - Low Accidents - ADS A Initiation Drywell Low Low, Level 1 Pressure - High
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| - ADS A Initiation Reactor (2 channels arranged Vessel Water Level - Low in a 2-out-of-2 logic for Low Low, Level 1 B trip system) - ADS B Initiation Drywell Pressure - High
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| : 2) Manual ADS Initiation 5.b. Drywell Pressure 15.6.5 Loss of Coolant 1) Automatic Initiation DBA
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| - High Accidents - ADS A Initiation Drywell Pressure - High (2 channels arranged - ADS A Initiation Reactor in a 2-out-of-2 logic for Vessel Water Level - Low B trip system) Low Low, Level 1
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| - ADS B Initiation Reactor Vessel Water Level - Low Low Low, Level 1
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| : 2) Manual ADS Initiation 5.c. ADS Initiation 15.6.5 Loss of Coolant 1) Automatic Initiation DBA Timer Accidents - ADS A Trip System
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| : 2) Manual ADS Initiation (single channel logic)
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| GNRO-2023/00014 Page 26 of 44 Table 4 - ECCS Instrumentation Redundancy/Diversity (TS Table 3.3.5.1-1)
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| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation1 5.d. Reactor Vessel 15.6.5 Loss of Coolant 1) Automatic Initiation DBA Water Level - Low Accidents - ADS A Trip System Level 3 (Confirmatory) 2) Manual ADS Initiation (single channel logic) 5.e. LPCI Pumps B & 15.6.5 Loss of Coolant 1) Automatic Initiation DBA C Discharge Pressure Accidents - LPCI Pump B Discharge
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| - High Pressure - High
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| - LPCI Pump C Discharge (2-out-of-2 channels Pressure - High on one pump) - ADS A Trip System
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| : 2) Manual ADS Initiation 5.f. ADS Bypass Timer 15.6.5 Loss of Coolant 1) Automatic Initiation DBA (High Drywell Accidents - ADS A Trip System Pressure) 2) Manual ADS Initiation (2-out-of-2 channels) 5.g. Manual Initiation None None 1) Automatic Initiation N/A
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| - ADS A Initiation Reactor (2 push buttons per Vessel Water Level - Low ADS B trip system in a Low Low, Level 1 2-out-of-2 logic) - ADS B Initiation Reactor Vessel Water Level - Low Low Low, Level 1 Note 1: Redundant instrument operable train within the function are not listed due to high diversity for each function. The diverse functions are in fact redundant divisions. The following functions have redundant instrument channels within the function: 1.a, 1.b, 1.d, 2.a, 2.b, 2.d, 3.a, 3.b, 3.c, and 3.d.
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| GNRO-2023/00014 Page 27 of 44 A-5. Reactor Core Isolation Cooling (RCIC) System Instrumentation - TS Section 3.3.5.3 The purpose of the RCIC System instrumentation is to initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink and normal coolant makeup flow from the feedwater system is unavailable, such that initiation of the low pressure ECCS pumps does not occur.
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| The RCIC System Instrumentation design creates defense-in-depth from the redundancy of the channels for each trip system.
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| Diverse RCIC system instrumentation inputs (TS Table 3.3.5.3-1):
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| * 1. Reactor Vessel Water Level - Low Low, Level 2:
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| o 4 instrument channels arranged in a 1 of 2 taken twice logic.
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| * 3. Condensate Storage Tank Level - Low:
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| o 2 instrument channels arranged in a one-out-of-two logic.
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| * 4. Suppression Pool Water Level - High:
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| o 2 instrument channels arranged in a one-out-of-two logic.
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| Note: Functions 2 (RCIC - Reactor Vessel Water Level - High Trip) and 5 (RCIC -
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| Manual Initiation) are not included in the RICT program.
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| GNRO-2023/00014 Page 28 of 44 Table 5 - RCIC Instrumentation Redundancy/Diversity (Table 3.3.5.3-1)
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| UFSAR Redundant/Diverse Function Transient Accident Event Section Instrumentation1
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| : 1. Reactor Vessel 15.1.1 Loss of Feedwater 1) Automatic Initiation IMF-T Water Level - Low Heating, automatic flow - RCIC Initiation Reactor Low, Level 2 control Vessel Water Level -
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| 15.1.2 Feedwater Controller Low Low, Level 2 (4 channels, 2 per - HPCS Initiation Reactor IMF -T Failure - Max. Demand trip system, Vessel Water Level -
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| arranged in a 1- 15.1.3 Pressure Controller Low Low, Level 2 IMF -T out-of-2 taken Failure - Open - LPCI A & LPCS twice logic) 15.2.1 Pressure Controller Initiation Reactor IMF -T Failure - Closed Vessel Water Level -
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| 15.2.3 Turbine Trip Low Low Low, Level 1
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| - LPCI B & C Initiation IMF -T 15.2.4 MSIV - Closure Reactor Vessel Water IMF -T Level - Low Low Low, 15.2.5 Loss of Condenser Level 1 IMF -T Vacuum 2) Manual Initiation 15.2.6 Loss of AC Power IMF -T 15.2.7 Loss of Feedwater IMF -T Flow 15.3.1 Recirculation Pump IMF -T Trip 15.3.3 Recirculation Pump DBA Seizure 15.3.4 Recirculation Pump DBA Shaft Break 15.2.8/ Feedwater Line Break DBA 15.6.6 Outside Containment
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| : 3. Condensate Implicitly assumed for accidents that 1) Automatic Initiation:
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| N/A Storage Tank Level initiate RCIC - Condensate Storage
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| - Low Tank Level - Low
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| - Suppression Pool Water (2 channels Level - High arranged in a 1- 2) Manual Alignment out-of-2 logic)
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| : 4. Suppression Implicitly assumed for accidents that 1) Automatic Initiation:
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| N/A Pool Water Level - initiate RCIC - Condensate Storage High Tank Level - Low
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| - Suppression Pool Water (2 channels Level - High arranged in a 1- 2) Manual Alignment out-of-2 logic)
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| Note 1: Instrumentation in italics is the affected function crediting redundant channels.
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| GNRO-2023/00014 Page 29 of 44 A-6. Primary Containment and Drywell Isolation Instrumentation - TS Section 3.3.6.1 The isolation instrumentation automatically initiates closure of appropriate primary containment isolation valves (PCIVs) and drywell isolation valves. The function of the PCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated DBAs.
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| The primary containment and drywell isolation (PCDI) instrumentation design creates defense-in-depth from the redundancy of the channels for each trip system. Except for MSIVs, one isolation system is associated with the inboard primary containment and drywell isolation valves and the other isolation system is associated with the outboard primary containment and drywell isolation valves with the success criteria being closure of one of the two isolation valves. Each MSIV has a solenoid-operated pilot valve associated with each trip system. Both trip systems must be tripped to de-energize both solenoid-operated pilots on each MSIV to close the valve. Each trip system has two logic channels monitoring instrumentation for each of the isolation parameters. A trip in either logic channel will trip the trip system. This results in a one-out-of-two, taken twice logic.
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| Diverse primary containment and drywell isolation inputs (TS Table 3.3.6.1-1):
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| * 1. Main Steam Line Isolation o 1.a. Reactor Vessel Water Level - Low Low Low, Level 1
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| : 1. b. Main Steam Line Pressure - Low 1.d. Condenser Vacuum - Low 1.e. Main Steam Tunnel Ambient Temperature - High 4 channels arranged in a one-out-of-two taken twice logic to isolate the MSIVs.
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| 4 channels arranged in two 2-out-of-2 trip systems to isolate all MSIV drain lines. One 2-out-of-2 trip system is associated with the inboard valves and the other 2-out-of-2 trip system is associated with the outboard valves.
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| o 1.c. Main Steam Line Flow - High:
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| 16 channels, 4 channels for each steam line (2 per trip system) arranged in a 1-out-of-2 taken twice logic for the affected steam line to initiate isolation of the MSIVs.
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| o 1.f. Manual Initiation:
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| 4 channels, 1 per each push button. 2 push buttons input into one trip system and the other 2 push buttons input into the other trip system. Both
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| GNRO-2023/00014 Page 30 of 44 trip systems are arranged in a one-out-of-two logic for each trip system.
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| To close all MSIVs, both trip systems must actuate. To close the MSL drain valves, both channels in a trip system must actuate to close the inboard or outboard valves.
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| * 2. Primary Containment and Drywell Isolation (Note: Functions 2.c, 2.d, 2.e, and 2.f are not included) o 2.a. Reactor Vessel Water Level - Low Low, Level 2:
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| 4 channels arranged in a 2-out-of-2 logic per trip system.
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| o 2.b. Drywell Pressure - High:
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| 4 channels arranged in a 2-out-of-2 logic per trip system.
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| o 2.g. Containment and Drywell Ventilation Exhaust Radiation- High:
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| 4 channels arranged in a 2 of 4 channel logic per trip system.
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| o 2.h. Manual Initiation:
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| 4 pushbuttons (4 channels), two per trip system: The logic arrangement is isolation group specific, either 1 of 1 or 2 of 2 in either trip system based on the isolation group design.
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| * 3. Reactor Core Isolation Cooling (RCIC) System Isolation o 3.a. RCIC Steam Line Flow - High
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| : 3. b. RCIC Steam Line Flow - Time Delay 3.c. RCIC Steam Supply Line Pressure - Low 3.e. RCIC Equipment Room Ambient Temperature - High 3.f. Main Steam Line Tunnel Ambient Temperature - High 3.g. Main Steam Line Tunnel Temperature Timer 3.h. RHR Equipment Room Ambient Temperature - High 3.i. RCIC/RHR Steam Line Flow - High 3.j. Drywell Pressure - High:
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| 2 channels, 1 per trip system, arranged in a single channel logic for either the inboard or outboard isolation valve.
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| o 3.d. RCIC Turbine Exhaust Diaphragm Pressure - High:
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| 4 channels, 2 per trip system, arranged in a 2-out-of-2 logic for either the inboard or outboard isolation valve.
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| o 3.k. Manual Initiation: 1 channel which initiates a single trip system.
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| * 4. RWCU System Isolation o 4.a. Differential Flow - High
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| : 4. b. Differential Flow - Timer 4.c. RWCU Heat Exchanger Equipment Room Temperature - High 4.e. RWCU Heat Exchanger Valve Nest Area Temperature - High 4.f. Main Steam Line Tunnel Ambient Temperature - High
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| GNRO-2023/00014 Page 31 of 44 4.h. Standby Liquid Control System Initiation:
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| 2 channels, 1 per trip system, arranged in a single channel logic for either the inboard or outboard isolation valve.
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| o 4.d. RWCU Pump Rooms Temperature - High 4 channels, 2 per room, 1 per trip system per room, arranged in a single channel logic for either the inboard or outboard isolation valve.
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| o 4.g. Reactor Vessel Water Level - Low Low, Level 2:
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| 4 channels, 2 channels in each trip system arranged in a 2-out-of-2 logic for either inboard or outboard isolation valve.
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| o 4.i. Manual Initiation:
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| 4 pushbuttons, 1 per channel, arranged in a 2-out-of-2 logic for each trip system associated with the inboard or outboard isolation valves.
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| * 5. Residual Heat Removal (RHR) System Isolation o 5.a. RHR Equipment Room Ambient Temperature - High:
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| 4 channels, 2 per trip system and 2 per area arranged in one-out-of-two for either trip system.
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| o 5.b. Reactor Vessel Water Level - Low, Level 3
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| : 5. d. Drywell Pressure - High:
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| 4 channels, 2 per trip system arranged in a 2 of 2 logic for each trip system (inboard or outboard valve) o 5.c. Reactor Steam Dome Pressure - High 4 channels, 2 per trip system arranged in a 1 of 2 logic for either trip system (inboard or outboard valve) o 5.e. Manual Initiation:
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| 4 pushbuttons, 2 per trip system, arranged in a 2-out-of-2 logic for each isolation system (inboard or outboard valve).
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| GNRO-2023/00014 Page 32 of 44 Table 6 - PCDI Instrumentation Redundancy/Diversity (TS Table 3.3.6.1-1)
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| UFSAR Redundant/Diverse Function Transient Accident Event Section Instrumentation1
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| : 1. Main Steam Line Isolation 1.a. Reactor Vessel 15.2.7 Loss of Feedwater 1) Automatic Initiation IMF-T Water Level - Low Flow - Reactor Vessel Water Level Low Low, Level 1 15.6.5 Loss of Coolant - Low Low Low Level 1 DBA Accidents - Main Steam Line Pressure (4 channels, 2 per - Low system, arranged in a 15.2.8/ Feedwater Line 2) Manual Isolation DBA 1-out-of-2 taken twice 15.6.6 Break Outside logic) Containment 1.b. Main Steam Line 15.1.3 Pressure Controller 1) Automatic Initiation IMF-T Pressure - Low Failure - Open - Main Steam Line Pressure
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| - Low (4 channels, 2 per - Reactor Vessel Water Level system, arranged in a - Low Low Low, Level 1 1-out-of-2 taken twice 2) Manual Isolation logic) 15.6.4 Steam System 1) Automatic Initiation DBA Piping Break - Main Steam Line Pressure Outside - Low Containment - Reactor Vessel Water Level
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| - Low Low Low Level 1
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| - Main Steam Line Flow -
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| High
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| - Main Steam Tunnel Temperature - High
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| - Main Steam Tunnel Differential Temperature -
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| High
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| : 2) Manual Isolation 1.c. Main Steam Line 15.6.4 Steam System 1) Automatic Initiation DBA Flow - High Piping Break - Main Steam Line Flow -
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| Outside High (16 channels, 4 Containment - Reactor Vessel Water Level channels for each - Low Low Low Level 1 steam line (2 per trip - Main Steam Line Pressure system) arranged in a - Low 1-out-of-2 taken twice - Main Steam Tunnel logic for the affected Temperature - high steam line) - Main Steam Tunnel Differential Temperature -
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| High
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| : 2) Manual Isolation
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| GNRO-2023/00014 Page 33 of 44 Table 6 - PCDI Instrumentation Redundancy/Diversity (TS Table 3.3.6.1-1)
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| UFSAR Redundant/Diverse Function Transient Accident Event Section Instrumentation1 1.d. Condenser 15.2.5 Loss of condenser 1) Automatic Initiation IMF-T Vacuum - Low Vacuum - Condenser Vacuum - Low
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| - Reactor Vessel Water Level (4 channels, 2 per - Low Low Low, Level 1 system, arranged in a 2) Manual Isolation 1-out-of-2 twice logic) 1.e. Main Steam 15.6.4 Steam System 1) Automatic Initiation DBA Tunnel temperature - Piping Break - Main Steam Line Flow -
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| High Outside High Containment - Reactor Vessel Water Level (4 channels, 2 per - Low Low Low Level 1 system, arranged in a - Main Steam Line Pressure 1-out-of-2 twice logic) - Low
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| - Main Steam Tunnel Temperature - High
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| - Main Steam Tunnel Differential Temperature -
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| High
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| : 2) Manual Isolation 1.f. Manual Initiation None None 1) Automatic Initiation N/A
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| - See Functions 1.a through (4 channels, 2 per 1.e.
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| system, arranged in a 1-out-of 2 twice logic for MSIVs or 2-out-of-2 for MSIV drains)
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| : 2. Primary Containment and Drywell Isolation 2.a. Reactor Vessel 15.6.5 Loss of Coolant 1) Automatic Initiation DBA Water Level - Low Accidents - Reactor Vessel Water Level Low, Level 2 - Low Low, Level 2
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| - Drywell Pressure - High (4 channels, arranged combined with in a 2-out-of-2 logic per - RWCU Isolation Signals for trip system) all Groups
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| : 2) Manual Isolation 2.b. Drywell Pressure - 15.6.5 Loss of Coolant 1) Automatic Initiation DBA High Accidents - Reactor Vessel Water Level
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| - Low Low, Level 2 (4 channels arranged - Drywell Pressure - High in a 2-out-of-2 logic per 2) Manual Isolation trip system)
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| GNRO-2023/00014 Page 34 of 44 Table 6 - PCDI Instrumentation Redundancy/Diversity (TS Table 3.3.6.1-1)
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| UFSAR Redundant/Diverse Function Transient Accident Event Section Instrumentation1 2.g. Containment and 15.6.5 Loss of Coolant 1) Automatic Initiation DBA Drywell Exhaust Accidents - Reactor Vessel Water Level Radiation - High - Low Low, Level 2
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| - Drywell Pressure - High (4 channels arranged - Containment Purge in a 2 of 4 channel Isolation Radiation - High logic per trip system) 2) Manual Isolation 2.d. Manual Initiation None None 1) Automatic Initiation N/A
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| - See Functions 2.a through (4 pushbuttons,4 2.c channels, two per trip system: The logic arrangement is isolation group specific, either 1 of 1 or 2 of 2 logic.)
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| GNRO-2023/00014 Page 35 of 44 Table 6 - PCDI Instrumentation Redundancy/Diversity (TS Table 3.3.6.1-1)
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| UFSAR Redundant/Diverse Function Transient Accident Event Section Instrumentation1
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| : 3. Reactor Core Isolation Cooling (RCIC) System Isolation 3.a. RCIC Steam Line These Functions are not 1) Automatic Initiation N/A Flow - High ** assumed in any UFSAR - RCIC Steam Line Flow -
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| 3.b. RCIC Steam Line transient or accident analysis, High & RCIC Steam Line N/A Flow - Time Delay ** since bounding analyses are Flow - Time Delay performed for large breaks - RCIC Steam Supply 3.c. RCIC Steam such as recirculation or MSL Pressure - Low *** N/A Supply Line Pressure - breaks. - RCIC Turbine Exhaust Low ** Diaphragm Pressure -
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| 3.d. RCIC Turbine RCIC isolations are credited in High *** N/A Exhaust Diaphragm UFSAR Chapter 3.6 for high - RCIC Equipment Room Pressure - High energy line leakage. Ambient Temperature -
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| 3.e. RCIC Equipment High N/A Room Ambient - Main Steam Line Tunnel Temperature - High &
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| Temperature - High **
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| Timer 3.f. Main Steam Line - RHR Equipment Room N/A Tunnel Ambient Ambient Temperature -
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| Temperature - High ** High 3.g. Main Steam Line - RCIC/RHR Steam Line N/A Tunnel Temperature Flow - High Timer **
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| : 2) Manual Isolation 3.h. RHR Equipment N/A Room Ambient Temperature - High **
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| *** Protective Trips 3.i. RCIC/RHR Steam N/A Line Flow - High **
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| 3.j. Drywell Pressure- Loss of Coolant Accident for 1)Automatic N/A High ** which RCIC is not credited. - RCIC Steam Pressure -
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| This is a protective trip. Low
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| : 2) Manual Isolation 3.k. Manual Initiation None None 1) Automatic Initiation N/A
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| - See Functions 3.a. thru 3.j.
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| (1 channel initiates a single trip system)
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| ** Functions with 2 channels, 1 per trip system in a single channel logic for either inboard or outboard isolation valve Functions with 4 channels, 2 per trip system in a 2-out-of-2 logic for either inboard or outboard isolation valve
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| GNRO-2023/00014 Page 36 of 44 Table 6 - PCDI Instrumentation Redundancy/Diversity (TS Table 3.3.6.1-1)
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| UFSAR Redundant/Diverse Function Transient Accident Event Section Instrumentation1
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| : 4. Reactor Water Cleanup (RWCU) System Isolation 4.a. Differential Flow - These Functions are not 1) Automatic Initiation N/A High ** assumed in any UFSAR - Differential Flow - High &
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| 4.b. Differential Flow - transient or accident analysis, Differential Flow - Time N/A Timer ** since bounding analyses are Delay performed for large breaks - RWCU Area (Functions such as MSLBs. 4.c.- 4.f) Temperature -
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| 4.c. RWCU Heat These Functions are not High N/A Exchanger Equipment assumed in any UFSAR - Reactor Vessel Level - Low Room Temperature - transient or accident analysis. Low, Level 2§ High ** 2) Manual Isolation 4.d. RWCU Pump Functions are credited in the N/A
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| § Reactor Vessel Level -Low Rooms Temperature - high energy line break Low Level 2 will isolate High analysis described in UFSAR Sections 3.6 and UFSAR RWCU system leak if leak 4.e. RWCU Heat Appendix 3B. detection isolation systems N/A Exchanger Valve Nest fail.
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| Area Room Temperature - High **
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| 4.f. Main Steam Line These Functions are not N/A Tunnel Ambient assumed in any UFSAR Temperature - High ** transient or accident analysis, 4.g. Reactor Vessel since bounding analyses are N/A Water Level - Low performed for large breaks Low, Level 2
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| * such as MSLBs.
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| 4.h. Standby Liquid 15.8 Anticipated 1) Manual Isolation ATWS Control System transients Without Initiation ** SCRAM 4.i. Manual Initiation ** None None 1) Automatic Initiation N/A See Functions 4.a. thru 4.h.
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| ** Functions with 2 channels, 1 per trip system in a single channel logic for either inboard or outboard isolation valve Functions with 4 channels, 2 per room, 1 per trip system per room, arranged in a single channel logic for either the inboard or outboard isolation valve
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| ** Functions with 4 channels (or pushbuttons), 2 channels in each trip system arranged in a 2-out-of-2 logic for either inboard or outboard isolation valve
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| GNRO-2023/00014 Page 37 of 44 Table 6 - PCDI Instrumentation Redundancy/Diversity (TS Table 3.3.6.1-1)
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| UFSAR Redundant/Diverse Function Transient Accident Event Section Instrumentation1
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| : 5. RHR System Isolation 5.a. RHR Equipment This Function is not assumed 1) Automatic Initiation N/A Room Ambient in any UFSAR transient or - RHR Ambient Temperature Temperature - High accident analysis, since - High
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| ** bounding analyses are - Reactor Vessel Water Level performed for large breaks Low - Level 3 such as MSLBs 2) Manual Isolation 5.b. Reactor Vessel Mitigation of vessel drain 1) Automatic Initiation N/A Water Level - Low, down, which is not assumed - Reactor Vessel Water Level Level 3 in any UFSAR transient or - Low, Level 3 accident analysis. - Drywell Pressure High
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| : 2) Manual Isolation 5.c. Reactor Steam This interlock is provided only 1) Automatic Initiation N/A Dome Pressure - for equipment protection to - Reactor Steam Dome High
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| * prevent an intersystem LOCA Pressure - High
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| * scenario and credit for the - Reactor Vessel Water Level interlock is not assumed in Low, Level 3 the accident or transient 2) Manual Isolation analysis in the UFSAR.
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| 5.d. Drywell Pressure - This Function is not assumed 1) Automatic Initiation N/A High in any UFSAR transient or - Reactor Vessel Water Level accident analysis, since - Low, Level 3 bounding analyses are - Drywell Pressure High performed for large breaks 2) Manual Isolation such as MSLBs 5.e. Manual Initiation None None 1) Automatic Initiation N/A
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| - See Functions 5.a thru 5.e.
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| Functions with 4 channels, 2 per trip system and 2 per area arranged in 1-out-of-2 logic for
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| ** either trip system (inboard or outboard isolation valve)
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| Functions with 4 channels (or pushbuttons), 2 per trip system arranged in a 2-out-of-2 logic for each trip system (inboard or outboard isolation valve)
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| ** Function with 4 channels, 2 per trip system arranged in a 1 of 2 logic for either trip system (inboard or outboard isolation valve)
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| Note 1: Instrumentation in italics is the affected function crediting redundant channels.
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| GNRO-2023/00014 Page 38 of 44 A-7. Residual Heat Removal (RHR) Containment Spray System Instrumentation -
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| TS Section 3.3.6.3 The RHR Containment Spray System is an operating mode of the RHR System that is initiated to condense steam in the containment atmosphere. This ensures that containment pressure is maintained within its limits following a LOCA.
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| The RHR containment spray (CS) instrumentation design creates defense-in-depth from the redundancy of the channels for each trip system. Each or the two Spray Trains is initiated by the division specific control logic.
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| Diverse RHR CS system instrumentation inputs (TS Table 3.3.6.3-1):
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| * 1. Drywell Pressure - High:
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| o 4 instrument channels arranged in a 1 of 2 logic on either trip system.
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| * 2. Containment Pressure - High:
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| o 4 instrument channels arranged in a 1 of 2 logic on either trip system.
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| * 3. Reactor Vessel Water Level - Low Low Low, Level 1:
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| o 4 instrument channels arranged in a 1 of 2 taken on either trip system.
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| * 4. System A and System B Timers:
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| o 2 instrument channels arranged in a single channel logic for each trip system.
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| GNRO-2023/00014 Page 39 of 44 Table 7 - RHR Containment Spray Instrumentation Redundancy/Diversity (TS Table 3.3.6.3-1)
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| UFSAR Transient Redundant/Diverse Function Event Section Accident Instrumentation 1 RHR Containment Spray System Instrumentation
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| : 1. Drywell Pressure - 15.6.5 Loss of 1) Automatic Initiation DBA High Coolant - Drywell Pressure High Accidents - Containment Pressure -
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| (4 channels arranged High in a 1 of 2 logic on - Reactor Vessel Water Level either trip system) - Low Low Low, Level 1
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| : 2) Manual Initiation 2
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| : 2. Containment 15.6.5 Loss of 1) Automatic Initiation DBA Pressure - High Coolant - Drywell Pressure High Accidents - Containment Pressure -
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| (4 channels arranged High in a 1 of 2 logic on - Reactor Vessel Water Level either trip system) - Low Low Low, Level 1
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| : 2) Manual Trip of Unit Coolers
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| : 3. Reactor Vessel 15.6.5 Loss of 1) Automatic Initiation DBA Water Level - Low Low Coolant - Drywell Pressure High Low, Level 1 Accidents - Containment Pressure -
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| High (4 channels arranged - Reactor Vessel Water Level in a 1 of 2 logic on - Low Low Low, Level 1 either trip system) 2) Manual Initiation 2
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| : 4. System A and Implicitly assumed for 1) Automatic Initiation DBA System B Timers accidents that initiate - Operable System A Timer Containment Sprays - Operable System B Timer (2 channels arranged 2) Manual Initiation 2 in a 1 of 1 logic on either trip system)
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| Notes:
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| 1: Instrumentation in italics is the affected function crediting redundant channels 2: Manual initiation is not an Instrumentation Tech. Spec. Function, but is addressed by LCO 3.6.1.7.A for RHR CS.
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| GNRO-2023/00014 Page 40 of 44 A-8. Relief and Low-Low Set (LLS) Instrumentation - TS Section 3.3.6.5 The safety relief valves (SRV) prevent overpressurization of the nuclear steam system.
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| Instrumentation is provided to support two modes of SRV operation - the relief function (all valves) and the LLS function (selected valves).
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| The SRV Relief and LLS Instrumentation design creates defense-in-depth from the redundancy of the channels for each trip system. The relief instrumentation consists of two trip systems, with each trip system actuating one solenoid for each SRV. There are two solenoids per SRV, each being able to open its SRV.
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| The relief mode is divided into three setpoint groups (the low with one SRV, the medium with ten SRVs, and the high with nine SRVs). The SRV relief function is actuated by 2-out-of-2 logic for each trip system of the three separate setpoint groups. There are three trip units (one per setpoint group) for each channel.
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| The LLS function is also divided into three setpoint groups (the low with one SRV, the medium with one SRV, and the high with four SRVs). The low and medium groups lower both the open and close setpoints. The high group only lowers the close setpoint. This design causes the six valves to remain open longer to reduce the potential for cycling.
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| The LLS logic is enabled and sealed in upon initial SRV actuation in any of the relief setpoint groups. The LLS is actuated by either 1-out-of-1 (i.e., single channel) logic (low, medium) or 2-out-of-2 logic (high).
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| Diverse Relief and LLS system instrumentation inputs (TS 3.3.6.5):
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| * Reactor Steam Dome Pressure - High:
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| o 4 instrument channels, 2 per trip system, with output to each trip system in each of three separate setpoint groups (high, medium, and low)
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| Relief: 2 of 2 of the same setpoint group on either trip system LLS Arm: 1 of any 3 relief setpoint groups taken twice LLS Control Group 1: Low (1 SRV) - 1 of 1 on either trip system.
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| LLS Control Group 2: Medium (1 SRV) - 1 of 1 on either trip system LLS Control Group 3: High (4 SRVs) - 2 of 2 on either trip system
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| GNRO-2023/00014 Page 41 of 44 Table 8 - Relief and LLS Instrumentation Redundancy/Diversity (TS 3.3.6.4)
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| UFSAR Redundant/Diverse Function Transient Accident Event Section Instrumentation Relief and Low-Low Set (LLS) Instrumentation
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| : a. Relief Mode 15.1.2 Feedwater Controller 1) Automatic Initiation IMF-T Failure - Maximum - Operable Relief Trip (See logic Demand System A description 15.1.3 Pressure Control Failure - Operable Relief Trip IMF-T above) - Open System B 15.2.2 Generator Load IMF-T Rejection 15.2.3 Turbine Trip IMF-T 15.2.4 MSIV Closure IMF-T 15.2.5 Loss of Condenser IMF-T Vacuum 15.2.6 Loss of AC power IMF-T 15.3.1 Recirculation Pump Trip IMF-T 15.3.2 Fast Closure of Two IMF-T Main Recirc. Valves 15.3.3 Recirculation Pump DBA Seizure 15.3.4 Recirculation Pump DBA Shaft Break 15.6.4 Steam System Piping DBA Break Outside Containment
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| : b. Low-Low Set 15.2.2 Generator Load 1) Automatic Initiation IMF-T Mode Rejection - Operable Relief and 15.2.3 Turbine Trip LLS Trip System A IMF-T (see logic - Operable Relief and above) 15.2.4 MSIV Closure LLS Trip System B IMF-T
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| GNRO-2023/00014 Page 42 of 44 A-9. Loss of Power (LOP) Instrumentation - TS Section 3.3.8.1 Successful operation of the required safety functions of the ECCS is dependent upon the availability of adequate power sources for energizing the various components such as pump motors, motor operated valves, and the associated control components. The LOP instrumentation monitors the 4.16 kV emergency buses.
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| The loss of power (LOP) instrumentation design creates defense-in-depth from separate Loss of Voltage or Degraded Voltage Initiation Functions.
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| Each 4.16 kV emergency bus has its own independent LOP instrumentation and associated trip logic. The voltage for Division 1, 2, and 3 buses is monitored at two levels which can be considered as two different undervoltage functions: loss of voltage and degraded voltage.
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| Each Division 1 and 2 Emergency Bus:
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| * Loss of Voltage Function is monitored by four undervoltage relays on the emergency bus. The outputs of these relays are arranged in a one-out-of-two taken twice logic configuration.
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| * Degraded Voltage Function is monitored by four undervoltage relays on the emergency bus. The outputs of these relays are arranged in a one-out-of-two taken twice logic configuration.
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| * Four channels of each 4.16 kV Emergency Bus Undervoltage (Loss of Voltage)
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| Function per associated emergency bus are only required to be operable when the associated DG is required to be operable. This ensures that no single instrument failure can preclude the DG function.
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| The Division 3 Emergency Bus:
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| * Loss of Voltage Function is monitored by four undervoltage relays on the emergency bus. The outputs of these relays are arranged in a one-out-of-two taken twice logic configuration.
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| * Degraded Voltage Function is monitored by four undervoltage relays on the emergency bus. The outputs of these relays are arranged in a one-out-of-two taken twice logic configuration.
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| Four channels of each 4.16 kV Emergency Bus Undervoltage (Loss of Voltage)
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| Function per associated emergency bus are only required to be operable when the associated DG is required to be operable. This ensures that no single instrument failure can preclude the DG function.
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| GNRO-2023/00014 Page 43 of 44 Regulatory Guide 1.174 Revision 3 - Section 2.1.1 Defense-in-Depth In accordance with the principles contained within 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, defense-in-depth consist of several elements and consistency with the defense-in-depth philosophy is maintained if the following occurs:
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| * A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation.
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| o The GGNS TS reflects this balance by allowing one 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.
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| * Over-reliance on programmatic activities as compensatory measures associated with the change in the licensing basis is avoided.
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| o 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.
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| * System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g., no risk outliers) o System redundancy, independence, and diversity remain the same as in the as designed condition. The number of operable functions has not been decreased, 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.
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| * Defense against potential common-cause failures is preserved, and the potential for the introduction of new common-cause failure mechanisms is assessed.
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| o This LAR does not impact the original determination of common-cause failure for the applicable instrumentation and its functions. It may allow the completion times (CTs) to be extended for one or two channels in a function and be inoperable prior to placing the channel in trip. Placing the channel in trip fulfills the channels trip function to perform the safety function of the applicable system.
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| * Independence of barriers is not degraded.
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| o Barriers are not affected by this LAR.
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| GNRO-2023/00014 Page 44 of 44
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| * Defenses against human errors are preserved.
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| o In the conditions listed in the TS, a potential extension of the TS CTs does not change any personnel actions required when the TS Action is entered.
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| Therefore, no change to the possibility of a human error is introduced and no change to the defenses against potential human error have been made.
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| * The intent of the plants design criteria is maintained.
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| o The design criteria of the applicable systems are maintained as reflected in the UFSAR. Redundancy, diversity of signal, and independence of trip/actuation channel functions are maintained with the requested change.
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| The change requested in the LAR does not physically change the applicable systems in any way. It only allows additional time, under certain low risk conditions in accordance with the RICT Program, to perform actions that the NRC has previously determined to be acceptable.
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| The defense-in-depth principles prescribed in Regulatory Guide 1.174, Revision 3 are met.
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| Attachment 6 GNRO-2023/00014 GGNS RICT Program PRA Implementation Items (2 pages follow)
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| GNRO-2023/00014 Page 1 of 2 RICT Program PRA Implementation Items
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| ==1.0 INTRODUCTION==
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| The table below identifies the items that are required to be complete prior to implementation of the Risk-Informed Completion Time (RICT) Program at Grand Gulf Nuclear Station Unit 1 (GGNS). All issues identified below will be addressed and any associated changes will be made and reflected in the PRA of record prior to implementation of the RICT Program.
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| Table A6-1: RICT Program PRA Implementation Items No. Description Implementation Item Current RPS logic models five 1 RPS logic will be rearranged to credit the SCRAM functions and credits these appropriate instrumentation functions for each functions for all initiating events.
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| modeled initiating event. This is a matter of These signals are:
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| model maintenance and does not introduce
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| * High reactor vessel pressure new methods.
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| * Low reactor vessel water level
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| * High reactor vessel water level
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| * Turbine Control/Stop Valves
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| * SDV Level Crediting all five functions for all initiating events is non-conservative because the functions are designed to respond to specific initiating events, and no initiating event has been identified that challenges all five functions. This implementation item is related to LCO 3.3.1.1, Condition A.
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| 2 The PRA only models PCIV A pre-existing large containment isolation instrumentation and valves identified failure will be added as a conservative as potential large early release surrogate for unmodeled PCIV instrumentation pathways. An appropriate and and pathways. This is a matter of model conservative surrogate is needed to maintenance and does not introduce new model unavailability of unmodeled methods.
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| PCIV instrumentation and valves included in this application. This implementation item is related to LCOs 3.3.6.1, Condition A; 3.6.1.2, Condition C; 3.6.1.3, Condition A; and 3.6.5.3 Condition A.
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| GNRO-2023/00014 Page 2 of 2 No. Description Implementation Item 3 The PRA does not model all As a surrogate for any unmodeled ADS or instrumentation functions associated Relief instrumentation unavailability, the division with ADS instrumentation and LLS specific safety relief valve (SRV) solenoid pilot Relief instrumentation associated with valves will be used as surrogates. GGNS will this application. For those unmodeled add the pilot solenoid valve failure logic to the functions, a conservative surrogate SRV models including division specific will be added to the model. This supports. This is a matter of model implementation item is related to LCO maintenance and does not introduce new 3.3.5.1, Condition F (functions 4.d, methods.
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| 5.d) and Condition G (functions 4.h, 5.g), and LCO 3.3.6.5, Condition A.
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| ENCLOSURE 1 GRAND GULF NUCLEAR STATION License Amendment Request Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF lnitiative4b" LIST OF REVISED REQUIRED ACTIONS TO CORRESPONDING PRA FUNCTIONS (54 Pages to Follow)
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| GNRO-2023/00014 Page 1 of 54 List of Revised Required Actions to Corresponding PRA Functions
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| ==1.0 INTRODUCTION==
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| Section 4.0, "Limitations and Conditions", Item 2 of the NRC Final Safety Evaluation (Reference 1) for Nuclear Energy Institute (NEI) Topical Report NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines", Revision O (Reference 2), identifies the following needed content:
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| * The license amendment request (LAR) will provide identification of the TS Limiting Conditions for Operation (LCOs) and action requirements to which the RMTS will apply.
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| * The LAR will provide a comparison of the TS functions to the PRA modeled functions of the structures, systems, and components (SSCs) subject to those LCO actions.
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| * 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.
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| This enclosure provides confirmation that the Grand Gulf Nuclear Station (GGNS) 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 for Section 4.0, Item 2 of the NRC Final Safety Evaluation. The scope of the comparison includes each of the TS LCO conditions and associated required actions within the scope of the RICT Program.
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| 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:
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| * Column "Tech Spec Description": Lists all of the LCOs and condition statements within the scope of the RICT Program .
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| * Column "SSCs Covered by TS LCO Condition and Applicable Mode(s)": List the SSCs addressed by each action requirement. Note that SSCs not applicable to the GGNS RICT Program are not listed.
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| * Column "Modeled in PRA?": Indicates whether the SSCs addressed by the TS LCO Condition are included in the PRA.
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| E1 - 1
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| GNRO-2023/00014 Page 2 of 54
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| * Column "Function Covered by TS LCO Condition": Lists a summary of the required functions from the design basis analyses.
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| * Column "Design Success Criteria": Provides a summary of the success criteria from the design basis analyses.
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| * Column "PRA Success Criteria": List the function success criteria modeled in the PRA.
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| * Column "Comments": Provides the justification or resolution to address any inconsistencies between the TS and PRA functions regarding the scope of SSCs and the success 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-A, Revision 0.
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| 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 the table. Scope differences are justified by identifying appropriate surrogate events which permit a risk evaluation to be completed using the Configuration Risk Management Program tool for the RICT Program. Differences in success criteria typically arise due to the requirement in the American Society of Mechanical Engineers (ASME) / American Nuclear Society (ANS) RA-Sa-2009 PRA Standard (hereafter "ASME/ANS PRA Standard") (Reference 3) 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 ASME/ANS PRA standard as required by NEI 06-09-A, Revision 0.
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| Examples of calculated RICT 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). These example calculations demonstrate the scope of the SSCs covered by TSs modeled in the PRA. Note that the more limiting of the core damage frequency (CDF) and large early release frequency (LERF) RICT result is shown.
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| Following implementation of the RICT Program, the actual RICT values will be calculated using the actual plant configuration and the current revision of the PRA model representing the as-built, as-operated condition of the plant, as required by NEI 06-09-A and the NRC Final Safety Evaluation. The actual RICT values may differ from the RICTs presented in this enclosure.
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| E1 - 2
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| GNRO-2023/00014 Page 3 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.1 .7.D One SLC Two SLC Yes Provide the One SLC subsystem Same as design SSCs are modeled subsystem subsystems capability for success criteria consistent with the TS inoperable for bringing the scope and can be directly reasons other than reactor from included in the Electronic Condition A, B, or full power to a Risk Assessment Tool C. subcritical, (ERAT) for the RICT xenon free program.
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| shutdown without taking credit for control rod movement 3.3.1.1.A Reactor Protection 1. Intermediate Range Monitors (IRM)
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| System (RPS) 1a. Neutron Flux - No Reactor Trip One Neutron Flux - None Failure of the RPS instrumentation High Initiation High channel in each mechanical seal-in-relays (Eight channels , two (SCRAM) RPS trip system to transfer was chosen as
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| - One or more channels per RPS a conservative surrogate.
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| required channels logic subsystem) One relay from either inoperable 1b. lnop No SCRAM One lnop. Channel in None group in a system will be each RPS trip system marked unavailable for a (Eight channels , two channel out of service channels per RPS logic subsystem) (OOS) in the RICT (e.g.,
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| K14A and K14E). This is considered conservative because it bypasses other RPS instrumentation (RPS level, pressure, etc).
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| E1 - 3
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| GNRO-2023/00014 Page 4 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.1.1.A 2. Average Power Range Monitors (APRM)
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| (cont.) 2a . Neutron Flux - No SCRAM One Neutron Flux - None Failure of the RPS High, Setdown High (Setdown) mechanical seal in relays (Four APRM channel in each RPS to transfer was chosen as channels and four trip system a conservative surrogate.
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| Voter channels) Both relays in the affected 2b. Fixed Neutron No SCRAM One Fixed Neutron Flux None subsystem will be marked Flux- High - High channel in each unavailable for a channel (Four APRM RPS trip system OOS in the RICT (e.g.,
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| channels and four K14A and K14E). This is considered conservative Voter channels) because it bypasses other 2c. lnop No SCRAM One lnop channel in None RPS instrumentation (Four APRM each RPS trip system (RPS level, pressure, etc).
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| channels and four Voter channels) 2d . Flow Biased No SCRAM One Flow Biased None Simulated Thermal Simulated Thermal Power - High (Four Power - High channel APRM channels and in each RPS trip four Voter channels) system 2e . 2-Out-Of-4 Voter No SCRAM One channel in each None (Four channels) RPS trip system 2f. OPRM Upscale No SCRAM One channel in each None (FourOPRM RPS channels and four Voter channels)
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| E1 - 4
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| GNRO-2023/00014 Page 5 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.1.1.A 3. Reactor Vessel Yes SCRAM One Reactor Vessel Same as design SSCs are modeled (cont.) Steam Dome Steam Dome Pressure success criteria. consistent with the TS Pressure - High - High channel in each Logic will be scope and can be directly (Four channels) RPS trip system adjusted in ERAT included in the Electronic to limit credit to Risk Assessment Tool applicable (ERAT) for the RICT initiators. program.
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| (See Note 10)
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| : 4. Reactor Vessel Yes SCRAM One RVWL - Low Same as design Water Level (RVWL) channel in each RPS success criteria.
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| - Low , Level 3 trip system Logic will be (Four channels) adjusted in ERAT to limit credit to applicable initiators.
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| (See Note 10)
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| : 5. Reactor Vessel Yes SCRAM One RVWL - High Same as design Water Level (RVWL) channel in each RPS success criteria.
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| - High, Level 8 trip system Logic will be (Four channels) adjusted in ERAT to limit credit to applicable initiators.
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| (See Note 10)
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| : 6. Main Steam No SCRAM Four MSIVs - Closure None Failure of the RPS Isolation Valve channels, two in each trip mechanical seal-in-relays (MSIV) - Closure system in three steam was chosen as a (Sixteen channels , lines conservative surrogate.
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| two per MSIV, one Both relays in the affected from each trip subsystem will be marked system) unavailable for a channel E1 - 5
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| GNRO-2023/00014 Page 6 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.1.1.A 7. Drywell Pressure - No SCRAM One Drywell Pressure - None OOS in the RICT (e.g.,
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| (cont.) High High channel in each of K14A and K14E). This is (Four channels) two trip systems conservative because it does not credit diverse RPS instrumentation (RPS level, pressure, etc).
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| : 8. Scram Discharge Volume Water Level - High 8a . Transmitter/Trip Yes SCRAM One Transmitter/ Trip Same as design SSCs are modeled Unit Unit ch annel in each success criteria. consistent with the TS (Four channels) RPS trip system Logic will be scope and can be directly adjusted in ERAT included in the Electronic to limit credit to Risk Assessment Tool applicable (ERAT) for the RICT initiators. program.
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| (See Note 10) 8b. Float Switch Yes SCRAM One Transmitter/ Level Same as design (Four channels) switch - High channel in success criteria.
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| each RPS trip system Logic will be adjusted in ERAl to limit credit to applicable initiators.
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| (See Note 10)
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| E1 - 6
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| GNRO-2023/00014 Page 7 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.1.1.A 9. Turbine Stop Valve Yes SCRAM Four TSV - Closure Same as design SSCs are modeled (cont.) (TSV) Closure Trip channels , two in each success criteria. consistent with the TS Oil Pressure - Low trip system in three of Logic will be scope and can be directly (Eight channels, two four TSVs adjusted in ERAT included in the Electronic TSV channels per to limit credit to Risk Assessment Tool RPS logic applicable (ERAT) for the RICT subsystem) initiators. program.
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| (See Note 10)
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| : 10. Turbine Control Yes SCRAM One TCV Fast Same as design Valve (TCV) Fast Closure Trip Oil success criteria.
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| Closure, Trip Oil Pressure - Low Logic will be Pressure - Low channel in each adjusted in ERAT (Four channels) RPS trip system to limit credit to applicable initiators.
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| (See Note 10) 11 . Reactor Mode No SCRAM One Reactor Mode None Failure of the RPS Switch - Shutdown Switch- Shutdown mechanical seal-in-relays Position Position channel in was chosen as a (Four channels) each RPS trip system conservative surrogate.
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| Both relays in the affected subsystem will be marked unavailable for a channel OOS in the RICT (e.g.,
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| K14A and K14E). This is considered conservative because it does not credit other RPS instrumentation (RPS level, pressure , etc).
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| E1 - 7
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| GNRO-2023/00014 Page 8 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition
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| : 12. Manual Scram Not SCRAM One Manual Scram Operator Manual switch output (Four channels) explicitly channel in each RPS successfully relays will be used as an trip system scrams reactor equivalent surrogate.
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| 3.3.1.1.B One or more See 3.3.1.1.A Not SCRAM See 3.3.1.1.A Same as Failure of the RPS functions with one or functions above explicitly functions above . design success mechanical seal-in-relays more required If two or more criteria to transfer was chosen as channels inoperable inoperable (and not a conservative surrogate.
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| in both trip systems. tripped) or bypassed Both relays in the affected channels of a function subsystem will be marked result in a Loss of unavailable for a channel Function , a RICT OOS in the RICT (e.g.,
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| would not be entered. K14A and K14E). This is considered conservative because it bypasses other RPS instrumentation (RPS level, pressure, etc).
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| 3.3.4.1.A End of Cycle Function a.1 Turbine Yes Trip Both Two Turbine Trip Valve Same as design SSCs are modeled Recirculation Stop Valve (TSV) - Recirculation Closure channels in success criteria consistent with the TS Pump Trip (EOC- Closure , Trip Oil Pumps either trip system scope and can be directly RPT) Pressure - Low included in the ERAT for Instrumentation - (Four channels) OR the RICT program.
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| One or more Function a.2. Turbine Yes Trip Both Same as design required channels Control Valve (TCV) Recirculation Two Turbine Governor success criteria inoperable . - Fast Closure, Trip Pumps Valve Fast Closure Trip Oil Pressure-Low Oil Pressure - Low (Four channels) channels in either trip system E1 - 8
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| GNRO-2023/00014 Page 9 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.4.2.A Anticipated Transient Function a. Reactor Yes Trips Two RVVVL - Low, Low, Same as design SSCs are modeled Without SCRAM Vessel Water Level Recirculation Level 2 channels in one success criteria consistent with the TS Recirculation Pump (RVVVL) - Low Low Pump of two trip systems scope and can be directly Trip (ATWS-RPT) Level2 associated with included in the ERAT for Instrumentation - One (Four channels) the trip system OR the RICT program.
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| or more channels (See Note 1) inoperable Function b. Reactor Yes Two RV Pressure - High Vessel (RV) Pressure channels in one of two
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| -High trip systems (Four channels)
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| (See Note 1) 3.3 .5.1.B Emergency Core 1. ECCS Actuation Instrumentation for Low Pressure Coolant Injection-A (LPCI) and Low Pressure Core Cooling System Spray (LPCS) Subsystems (ECCS) Actuate both One RVWL Level 1 Same as design SSCs are modeled 1.a. Reactor Vessel Yes Instrumentation - As Water Level - Low LPCI A and channel success criteria consistent with the TS required by Required Low Low, Level 1 LPCS scope and can be directly Action A.1 and (Two channels) OR included in the ERAT for referenced in Tech - the RICT program.
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| Spec Table 3.3.5.1-1 . 1.b. Drywell Pressure Yes
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| -High One Drywell Pressure
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| - High channel from (Two channels)
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| Two Subsystems
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| : 2. ECCS Actuation Instrumentation for LPCI B and LPCI C Subsystems 2.a. Reactor Vessel Yes Actuate both One RVWL - Level 1 Same as design SSCs are modeled Water Level - Low LPCI Band channel success criteria consistent with the TS Low Low, Level 1 LPCIC scope and can be directly (Two channels) OR included in the ERAT for 2.b. Drywell Pressure Yes the RICT program.
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| -High One Drywell Pressure
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| - High channel from (Two channels)
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| Two Subsystems E1 - 9
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| GNRO-2023/00014 Page 10 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3 .5.1.B 3. ECCS Actuation Instrumentation High Pressure Core Spray (HPCS) System (cont.) 3.a. Reactor Vessel Yes Actuate Two RVWL Level 2 Same as design SSCs are modeled Water Level - Low HPCS channels , one in each success criteria consistent with the TS Low, Level 2 subsystem scope and can be directly (Four channels) included in the ERAT for 3.b. Drywell Pressure Yes Actuate Two Drywell Pressure Same as design the RICT program.
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| -High HPCS - High channels , one success criteria (Four channels) in each subsystem 3.3.5.1.C ECCS 1. ECCS Actuation Instrumentation for LPCI A and LPCS Subsystems Instrumentation - As 1.c. LPCI Pump A Not Actuate LPCI One LPCI Pump A Same , but The actuation logic event required by Required Start - TD Relay Start TD Relay explicitly A system modeled using a for the LPCS and LPCIA Action A.1 and (One relay) division (See Note 9) single logic pumps will be used as a referenced in Tech failure event for surrogate. It is a Spec Table 3.3.5.1-1 . both pumps. conservative selection because it fails both pump start functions, whereas TD Relay fails only LPCI A only.
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| 1.d. Reactor Vessel Yes Permit LPCS One RVP-Low One RVP- Low The transmitters that Pressure (RVP) - orLPCIA pressure transmitter pressure supply input to the trip Low (Injection Injection supplying two transmitter (trip channels will be used as permissive) pressure trip unit units not conservative surrogates.
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| (Four pressure channels explicitly This is conservative ch annels) modeled) because no credit is given for redundancy in trip units.
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| E1 - 10
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| GNRO-2023/00014 Page 11 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.5.1.C 1.g. Manual Initiation Not Manually One PB for credited Manual action to Pushbutton failure (cont.) (One pushbutton explicitly actuate division start LP systems. bounded by human error (PB) for Division I) ECCS (See Note 9) event. Use of the human pumps error event to start LP systems is conservative because it fails manual start of all pumps versus one pump.
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| : 2. ECCS Actuation Instrumentation for LPCI B and LPCI C Subsystems 2.c. LPCI Pump B Not Actuate LPCI One LPCI Pump B Same , but The actuation logic event Start - TD Relay explicitly B system Start - TD Relay modeled using a for the LPCIB and LPCIC (One relay) division (See Note 9) single common pumps will be used as a cause logic surrogate. It is a failure event for conservative selection both pumps . because it fails both pump start functions, whereas TD Relay fails LPCI B only.
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| 2.d. Reactor Vessel Yes Permit LPCI One RVP-Low One RVP- Low The transmitters that Pressure - Low B or LPCI C pressure transmitter pressure supply input to the trip (Injection permissive) Injection supplyin g two transmitter (trip channels will be used as (Four pressure pressure trip unit units not conservative surrogates.
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| channels per channels explicitly This is conservative division) modeled) because no credit is given for redundancy in trip units.
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| E1 - 11
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| GNRO-2023/00014 Page 12 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.5.1.C 2.f. Manual Initiation Not Manually One PB for credited Manual action to Pushbutton failure (cont.) (One pushbutton explicitly actuate division start LP systems. bounded by human error (PB) for Division 11) ECCS (See Note 9) event. Use of the human pumps error event to start LP systems is conservative because it fails manual start of all pumps versus one pump.
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| : 3. ECCS Actuation Instrumentation HPCS System 3.c. Reactor Vessel Not Close HPCS One RVWL- High Level HPCS Level 8 Failure of RCIC Level 8 Water Level - High explicitly injection 8 channel logic card trip logic will be used as a Level8 valve to operation conservative surrogate for (Two channels) prevent interrupts HPCS instrument channels as it steam line flow to prevent is a direct failure of RCIC .
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| overflow. steam line flood. Condenser cooling will also be failed as a conservative surrogate.
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| (See Note 2) 3.h. Manual Initiation Not Manually One PB for credited Manual action to Pushbutton failure (One pushbutton explicitly Actuate division start HPCS bounded by human error (PB)) HPCS pump (See Note 9) event in the PRA. Use of the human error event to start HPCS is equivalent to a logical OR to pushbutton circuit failure.
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| E1 - 12
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| GNRO-2023/00014 Page 13 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.5.1.D ECCS 3. ECCS Actuation Instrumentation HPCS System Instrumentation - As 3.d. Condensate Yes Automaticall One of two channels One of four SSCs are modeled required by Required y align of CST Level - Low channels of CST consistent with the TS Storage Tank (CST)
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| Action A.1 and Level - Low HPCS Level -Low scope and can be directly referenced in Tech (Two channels) suction from included in the ERAT for Spec Table 3.3.5.1-1 .
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| CST to the RICT program. Two Suppression additional non-TS CST-Pool for Level - Low channels are continued modeled explicitly with the HPCS same detail as TS operation OR channels .
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| 3.e. Suppression No None SPWL High channels are One of two channels-Pool Water Level not modeled because of SPWL - High (SPWL) - High operators are trained to (Two channels) maintain CST suction beyond the SPWL signal initiation. The CST level low channels will be used as conservative surro-gates crediting no diverse signal for suction swap.
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| 3.3 .5.1.E ECCS 1. ECCS Actuation Instrumentation for LPCI A and LPCS Subsystems Instrumentation - As 1.e. LPCS Pump No Open pump One low flow switch None The LPCS minimum flow required by Required minimum actuates valve is not modeled.
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| Discharge Flow -
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| Action A.1 and Low (Bypass) flow valve on (See Note 9) Taking the LPCS system referenced in Tech (One channel) low flow. OOS is used as a Spec Table 3.3.5.1-1 .
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| Drop signal conservative surrogate.
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| on not low flow.
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| E1 - 13
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| GNRO-2023/00014 Page 14 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.5.1.E 1.f. LPCI Pump A Yes Open pump One low flow switch Same as design SSCs are modeled (cont.) Discharge Flow - minimum actuates success criteria consistent with the TS Low (Bypass) flow valve on (See Note 9) scope and can be directly (One channel) low flow. included in the ERAT for Drop signal the RICT program.
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| on not low flow.
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| : 2. ECCS Actuation Instrumentation for LPCI B and LPCI C Subsystems 2.e. LPCI Pump B Yes Open pump One low flow switch Same as design SSCs are modeled and LPCI Pump C minimum actuates for success criteria consistent with the TS Discharge Flow - flow valve on associated pump scope and can be directly Low (Bypass) low flow. (See Note 9) included in the ERAT for (One channel per Drop signal the RICT program.
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| pump) on not low flow .
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| : 3. ECCS Actuation Instrumentation HPCS System 3.f. HPCS Pump Yes Open pump One high pressure Minimum flow SSCs are modeled Discharge Pressure minimum switch actuates valve opens consistent with the TS
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| - High (Bypass) flow valve on AND scope and can be directly (One channel) low flow One low flow switch included in the ERAT for AND high actuates. the RICT program.
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| 3.g. HPCS System Yes Flow Rate - Low discharge Drop of either signal (Bypass) pressure. closes bypass valve .
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| (See Note 9)
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| (One channel)
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| E1 - 14
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| GNR0-2023/00014 Page 15 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.5.1.F ECCS ADS initiation logic and instrumentation functions Instrumentation - Two RVWL - Low Low 4.a . Reactor Vessel Yes Initiate ADS Same as design SSCs are modeled As required by Water Level - Low Division 1 Level 1 channels in consistent with the TS success criteria Required Action A.1 Low Low , Level 1 either of two ADS scope and can be directly and referenced in (Two channels) actuation systems included in the ERAT for Tech Spec Table the RICT program.
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| 3.3.5.1-1.
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| 5.a . Reactor Vessel Yes Initiate ADS SSCs are modeled Water Level - Low Division 2 consistent with the TS Low Low , Level 1 scope and can be directly (Two channels) included in the ERAT for the RICT program.
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| 4.b . Drywell Yes Initiate ADS Two Drywell Pressure Same as design SSCs are modeled Pressure - High Division 1 - High channels in success criteria consistent with the TS (Two channels) either of two ADS scope and can be directly 5.b . Drywell Yes Initiate ADS actuation systems included in the ERAT for Pressure - High Division 2 the RICT program.
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| (Two channels) 4.d . Reactor Vessel Not ADS One RVWL - Low Same as design Division specific ADS Water Level - Low explicitly Permissive Level 3 channel in success criteria SRV pilot SOVs and Level3 Division 1 either of two ADS supports will be added to (Confirmatory) actuation systems the ADS model as (One channel) conservative surrogates 5.d . Reactor Vessel Not ADS for the LCO condition.
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| Water Level - Low explicitly Permissive This surrogate is Level3 Division 2 conservative as it takes (Confirmatory) no credit for other (One channel) functions.
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| E1 - 15
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| GNRO-2023/00014 Page 16 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3 .5.1.G ECCS ADS initiation logic and instrumentation functions Instrumentation - As 4.c. ADS Initiation Yes Initiate ADS One ADS Initiation Timer Same as design SSCs are modeled required by Required Timer consistent with the TS Timer Division channel on either of two success criteria Action A.1 and (One channel) 1 ADS actuation systems scope and can be directly referenced in Tech (See Note 9) included in the ERAT for Spec Table 3.3.5.1- 5.c. ADS Initiation Yes Initiate ADS Timer Timer Division the RICT program.
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| 1.
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| (One channel) 2 4.e. LPCS Pump Yes Permissive for Two pump discharge Same as design SSCs are modeled Discharge Pressure - ADS Actuation pressure - high success criteria consistent with the TS High Division 1 channels on one pump scope and can be directly (Two channels) in either ADS actuation included in the ERAT for 4.f. LPCI Pump A Yes system the RICT program.
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| Permissive for Discharge Pressure ADS Actuation
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| - High Division 1 (Two channels) 5.e . LPCI Pumps B Yes Permissive for
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| & C Discharge ADS Actuation Pressure - High Division 2 (Four channel) 4.g . ADS Bypass Yes Initiate ADS Two ADS Bypass Same as design SSCs are modeled Timer (High Drywell Bypass Timer Initiation Timer channel success criteria consistent with the TS Pressure) A if NOT High on either of two ADS scope and can be directly (Two Channels) Drywell actuation systems included in the ERAT for Pressure the RICT program.
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| 5.f. ADS Bypass Yes Initiate ADS Timer (High Drywell Bypass Timer Pressure) B if NOT High (Two Channels) Drywell Pressure E1 - 16
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| GNRO-2023/00014 Page 17 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3 .5.1.G 4.h . Manual Initiation Not Manual ADS Two ADS manual None Division specific ADS (cont.) (Two pushbuttons explicitly SRV Train A initiation PB in either of SRV pilot SOVs and (PB) per ADS trip Actuation two ADS actuation supports will be added to system) systems. the ADS model as 5.g . Manual Initiation Not Manual ADS conservative surrogates (Two pushbuttons explicitly SRV Train B for the LCO condition.
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| (PB) per ADS trip Actuation This surrogate is system) conservative as it takes no credit for other functions.
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| 3.3 .5.3.B Reactor Core 1. Reactor Vessel Yes RCIC initiation Two RVWL- Low Low Same as design SSCs are modeled Isolation Cooling Water Level - Low Level 2 channels (one in success criteria consistent with the TS (RCIC) System Low , Level 2 each division subsystem) scope and so can be Instrumentation - As channels directly included in the required by Required (Four channels) ERAT for the RICT Action A.1 and program.
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| referenced in Tech Spec Table 3.3.5 .3-1 E1 - 17
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| GNRO-2023/00014 Page 18 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3 .3.5.3.D RCIC System 3. Condensate Yes Initiate swap One Condensate Same as design SSCs are modeled Instrumentation - Storage Tank Level of RCIC Storage Tank Level success criteria consistent with the TS As required by -Low suction from - Low channel scope and can be directly Required Action (Two channels) CST to included in the ERAT for A.1 and Suppression the RICT program.
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| referenced in p,..,,..,, (")0 4 . Suppression Pool No None SPWL High channels are Tech Spec Table VIit: OUIJIJI e::;::;1u11 Water Level - High not modeled because 3.3.5.3-1 Pool Water Level -
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| (Two channels) operators are trained to High channel maintain CST suction beyond the SPWL signal initiation. The CST level low channels will be used as conservative surro-gates crediting no diverse signal for suction swap.
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| 3 .3.6.1.A Primary 1. Main Steam Line (MSL) Isolation Containment and 1.a. Reactor Yes Automatic Two RVWL Low Low Same as design SSCs are modeled Drywell Isolation Vessel Water isolation of Level 1 channels, success criteria consistent with the TS Instrumentation -
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| Level - Low Low Main Steam one in each trip scope and can be one or more Low , Level 1 Isolation system directly included in the required channels (Four channels) Valves ERAT for the RICT inoperable (MSIVs) program.
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| 1.b. Main Steam No Automatic Two MSL Pressure - None This signal is not Line Pressure - isolation of Low channels, one in modeled, so function Low MSIVs each trip system 1.a. channels will be (Four channels) used as surrogates for E1 - 18
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| GNRO-2023/00014 Page 19 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3 .3.6.1.A 1.c. Main Steam No Automatic Two MSL Flow - None this function. Any (cont .) Line Flow - High isolation of High channels in the inoperable subsystem (16 channels, four MSIVs effected MSL, one in channel will fail the per line) each trip system same subsystem channel of function 1.a 1.d. Condenser No Automatic Two Condenser None in the ERAT. This is Vacuum- Low isolation of Vacuum - Low , one conservative because (Four channels) MSIVs in each trip system diverse functions are not 1.e. Main Steam No Automatic Two Main Steam None credited. (See Note 7.a)
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| Tunnel Ambient isolation of Tunnel Temperature Temperature - MSIVs - High channels, one High in each trip system (Four channels) 1.f. Manual Not Manual Two PBs, one in Manual Pushbutton failures Initiation explicitly isolation of each trip system actuation of bounded by human error (Four pushbuttons MSIVs containment event in the PRA. Use of (PB)) isolation the human error event is conservative. (See Note 7.a)
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| : 2. Primary Containment and Drywell Isolation 2 .a. Reactor Not Automatic Two RVWL Low , Same as design PCIVs associated with Vessel Water explicitly isolation of Low , Level 2 success criteria the affected trip system Level - Low Low , PCIVs channels in either of will be used as a Level2 two trip systems conservative surrogate.
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| (Four channels) (See Note 7.b) 2 .b. Drywell Not Automatic Two Drywell Same as design Pressure - High explicitly isolation of Pressure High success criteria (Four channels) PCIVs channels in either of two trip systems E1 - 19
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| GNRO-2023/00014 Page 20 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.6.1.A 2.g. Containment No Automatic Two Containment Signal not PCIVs associated with (cont .) and Drywell isolation of and Drywell credited to close the affected trip system Ventilation PCIVs Ventilation Exhaust Purge Valves will be used as a Exhaust Radiation Radiation - High conservative surrogate.
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| - High channels in either of (See Note 7.b)
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| (Four channels) two trip systems 2.h. Manual Not Manual Either one or two Manual Initiation explicitly isolation of PBs in either trip actuation of (Four Pushbuttons PCIVs system containm ent (PB)) isolation
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| : 3. Reactor Core Isolation Cooling (RCIC) System Isolation 3.a. RCIC Steam No Automatic One Steam Line Flow - None RCIC steam line break is Line Flow - High Isolation of high channel on either not modeled. Failure of (Two channels) RCIC isolation system the RCIC will be used as a conservative surrogate.
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| 3.b. RCIC Steam No Automatic One Steam Line Flow - None Line Flow - Time Isolation of TD channel on either Delay RCIC isolation system (Two channels) 3.c. RCIC Steam No Automatic One RCIC Steam None Supply Line Isolation of Supply Pressure - Low Pressure - Low RCIC channels on either (Two channels) isolation system 3.d. RCIC Turbine No Automatic Two RCIC TEDP - High None Exhaust Diaphragm Isolation of channels on either Pressure (TEDP) - RCIC isolation system High (Four channels)
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| E1 - 20
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| GNRO-2023/00014 Page 21 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.6.1.A 3.e. RCIC No Automatic One RCIC ERAT - High None RCIC steam line break is (cont .) Equipment Room Isolation of channel on either not modeled. Failure of Ambient RCIC isolation system the RCIC will be used as Temperature a conservative surrogate.
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| (ERAT) - High (Two channels) 3.f. Main Steam No Automatic One MSL Tunnel None Line Tunnel Isolation of Ambient Temperature -
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| Ambient RCIC High on either isolation Temperature - High system (Two channels) 3.g. Main Steam No Automatic One MSL Tunnel None Line Tunnel Isolation of Temperature Timer on Temperature Timer RCIC either isolation system (Two channels) 3.h. RHR No Automatic One RHR ERAT - High None Equipment Room Isolation of channel on either Ambient RCIC isolation system Temperature (ERAT) - High (Four channels) 3.i. RCIC/RHR No Automatic One RCIC/RHR SLF - None Steam Line Flow Isolation of High channel on either (SLF) - High (Two RCIC isolation system channels) 3.j . Drywell No Automatic One Drywell Pressure - None RCIC steam line break is Pressure-High Isolation of High channel on either not modeled. Failure of (Two channels) RCIC isolation system the RCIC will be used as a conservative surrogate.
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| E1 - 21
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| GNRO-2023/00014 Page 22 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.6.1.A 3.k. Manual No Manual One Manual PB None RCIC steam line break is (cont .) Initiation (One Isolation of not modeled. Failure of Pushbutton) RCIC the RCIC will be used as a conservative surrogate.
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| : 4. Reactor Water Cleanup (RWCU) System Isolation 4.a. Differential No Automatic One Differential Flow - None RWCU isolation valves Flow - High (Two Isolation of High channel on either associated with the channels) RWCU isolation system affected trip system will valves be used as a conservative surrogate.
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| 4.b. Differential No Automatic One Differential Flow None (See Note 7. b)
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| Flow-Timer Isolation of Timer channel on either (Two channels) RCWU isolation system valves 4.c. RWCU Heat No Automatic One Heat Exch . Equip. None Exchanger Isolation of Room Temperature -
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| Equipment Room RCWU High channel on either Temperature - High valves isolation system (Two channels) 4.d. RWCU Pump No Automatic One RWCU Pump Room None Room Temperature Isolation of Temperature - High
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| - High RCWU channel on either (Four channels, two valves isolation system per area) 4.e. RWCU Heat No Automatic One RWCU Valve Nest None Exchanger Valve Isolation of Room Temperature -
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| Nest Area RCWU High channel on either Temperature - High valves isolation system (Two channels)
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| E1 - 22
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| GNRO-2023/00014 Page 23 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.6.1.A 4.f. Main Steam No Automatic One Main Steam Line None RWCU isolation valves (cont .) Line Tunnel Isolation of Tunnel Ambient associated with the Ambient RCWU Temperature - High on affected trip system will Temperature - High valves either isolation system be used as a (Two channels) conservative surrogate.
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| (See Note 7. b) 4.g. Reactor Vessel No Automatic Two RVWL Low , Low, None Water Level - Low Isolation of Level 2 channels on Low , Level 2 RCWU either isolation system (Four channels) valves 4.h. Standby Liquid No Automatic One SLC System None Control (SLC) Isolation of Initiation channel on System Initiation RCWU either isolation system (Two channels, one valves per SLC pump) 4.i. Manual Initiation Not Manual Two PBs on either Manual actuation (Four pushbuttons explicitly Isolation of isolation system of containment (PB)) RCWU isolation valves
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| : 5. Residual Heat Removal (RHR) System Isolation 5.a. RHR No Automatic One RHR Equipment None Relevant isolation valves Equipment Room Isolation of Room Ambient in these groups are Ambient RHR valves Temperature - High normally closed during Temperature - High channels on either power operations, and so (Four channels, two isolation system they are unmodeled.
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| per area) Therefore, a pre-existing large containment 5.b. Reactor Vessel No Automatic Two RVWL Low , Level 3 None isolation failure will be Water Level - Low, Isolation of channels on either Level3 RHR valves isolation systems used as a conservative (Four channels) surrogate. (See Note 7.c)
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| E1 - 23
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| GNRO-2023/00014 Page 24 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.6.1.A 5.c. Reactor Steam No Automatic One RSDP - High None Relevant isolation valves (cont .) Dome Pressure Isolation of channel on either in these groups are (RSDP) - High RHR valves isolation system normally closed during (Four channels) power operations, and so they are unmodeled.
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| 5.d. Drywell No Automatic Two Drywell Pressure None Therefore, a pre-existing Pressure - High Isolation of - High channels on large containment (Four channels) RHR valves either isolation system isolation failure will be 5.e. Manual No Manual Two PBs on either None used as a conservative Initiation isolation of isolation system surrogate. (See Note 7.c)
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| (Four pushbuttons RHR valves (PBs))
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| 3.3 .6.3.B Residual Heat 1. Drywell Pressure Yes Automatic One Drywell Pressure Same as design SSCs are modeled Removal (RHR) -High start of RHR - High channel on success criteria consistent with the TS Containment (Four channels) Contain ment either trip system scope and can be directly Spray System Spray included in the ERAT for Instrumentation - the RICT program As Required by
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| : 3. Reactor Vessel Yes Automatic One RVVVL - Low Same as design SSCs are modeled Required Action Water level (RVWL) start of RHR Low Low, Level 1 success criteria consistent with the TS A.1 and referenced
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| - Low Low, Level 1 Containment channel on either trip scope and can be directly in Tech Spec (Four channels) Spray system included in the ERAT for Table 3.3.6.3-1 the RICT program 3.3 .6.3.C Residual Heat 2. Containment Yes Automatic One Containment Same as design SSCs are modeled Removal (RHR) Pressure - High start of RHR Pressure - High success criteria consistent with the TS Containment (Four channels) Containment channel on either trip scope and can be Spray System Spray system directly included in the Instrumentation - ERAT for the RICT As Required by program.
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| E1 - 24
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| GNRO-2023/00014 Page 25 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.6.3.C Required Action 4. System A and Yes Delay auto One System Timer on Same as design SSCs are modeled (cont.) A.1 and referenced System B Timers initiation of either trip system success criteria consistent with the TS in Tech Spec (Two channels) RHR scope and can be Table 3.3.6 .3-1 Containment directly included in the Spray to ERAT for the RICT allow the program.
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| LPCI system to fulfill its function E1 - 25
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| GNRO-2023/00014 Page 26 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.6.5.A Relief and Low- Reactor Steam No Relief: Relief: Two RSDP None For the relief function, Low Set (LLS) Dome Pressure Prevent channels on either trip division specific SRV Instrumentation - (RSDP) vessel system pilot SOVs and supports one trip system instrumentation to overpressure LLS Arm: One RSDP will be added to the SRV inoperable support relief LLS : Lower channel on both trip relief model as function and LLS Relief subsystems (one in conservative surrogates function Setpoint on both trip systems) for the LCO condition.
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| (Two channels per three LLS Control : Two or This surrogate will have trip system) setpoint One RSDP channels an equivalent impact as groups in either trip system channel unavailability.
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| causing 1 specific to control (low) , 1 group For the LLS functions, the (medium) , or SRV Fail to Reclose 4 (high) failure mode probability group LLS will be doubled to account SRVSto for additional SRV cycles.
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| remain open This is conservative longer to because additional cycles prevent would not occur unless multiple both trip systems were actuations inoperable.
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| and postulated pressure loads on containment.
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| E1 - 26
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| GNRO-2023/00014 Page 27 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.8.1.A Loss of Power 1. Divisions I and II - 4.16 kV Emergency Bus Undervoltage (LOP) 1.a. Loss of Not Sense LOV Two LOV channels per Same, but A single event per division instrumentation -
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| Voltage (LOV) - explicitly on Division 1 bus modeled as a is used to model a loss of One or more 4 .16 kV Basis or 2 4.16 kV single logic emergency bus channels (Eight channels, emergency event for each undervoltage signal in the inoperable four per bus) bus and division PRA. This event is used as (Referenced in initiate power a surrogate for Tech Spec Table transfer to inoperability of any 3.3.8.1- 1)
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| DGs channel in these functions.
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| This is conservative 1.b. Loss of Not Time Delay One Time Delay Same, but Voltage - Time because diverse signals explicitly for power channel per logic modeled as a Delay subsystem single logic and redundant channels recovery are not explicitly credited (Four channels , event for each two per bus) division to start the standby DGs and transfer the 1.c. Degraded Not Sense DV on Two DV channel per Same, but emergency bus source.
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| Voltage (DV) - 4.16 explicitly Division 1 or bus modeled as a kV basis 2 4.16 kV single logic (Eight channels, four emergency event for each per bus) bus and division initiate power transfer to DGs 1.d. Degraded Not Time Delay One DV TD No-LOCA Same , but Voltage - Time Delay explicitly for power channel per logic modeled as a (Four channels , two recovery subsystem single logic per bus) event for each division E1 - 27
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| GNRO-2023/00014 Page 28 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.3.8.1.A 2. Division 3 - 4.16kV Emergency Bus Undervoltage (cont .)
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| 2 .a. Loss of Not Sense LOVon Two LOV channels Same , but Two events for Division 3 Voltage - 4.16 kV explicitly Division 3 modeled as two is used to model a loss of basis 4.16 kV logic events for Emergency Bus (Four channels) emergency the division Undervoltage Signal in the bus and PRA. This event is used initiate power as a surrogate for transfer to DG inoperability of any channel in these 2.b . Loss of Not Time Delay Two Time Delay Same, but Voltage - Time functions. This is explicitly for power channels modeled as two Delay (Two recovery logic events for conservative because channels) the division diverse signals and redundant channels are 2.c. Degraded Not Sense DV on Two DV channels Same as design not explicitly credited to Voltage - 4.16 kV explicitly Division 3 success criteria start the standby DGs and Basis 4.16 kV transfer the Emergency (Four channels) emergency Bus source .
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| bus and initiate power transfer to DG 2.d. Degraded Not Time Delay Two DVTime Delay, Same , but Voltage - Time explicitly for power No LOCA channels modeled as two Delay, No LOCA recovery logic events for (Two channels) the division 2.e. Degraded Not Time Delay Two DVTime Delay, Same , but Voltage - Time explicitly for power LOCA channels modeled as two Delay, LOCA recovery logic events for (Four channels) the division E1 - 28
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| GNRO-2023/00014 Page 29 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.5.1.A One low pressure Three LPCI trains Yes Low Two ECCS Low One ECCS Low SSCs are modeled ECCS and one LPC S pressure Pressure Pump Pressure Pump consistent with the TS injection/spray train injection/ subsystems subsystem scope and so can be subsystem (TS defines spray into directly included in the inoperable individual pump the Reactor ERAT for the RICT train as a Pressure program.
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| subsystem) Vessel (RPV) 3.5.1.8 High Pressure Core One HPCS train Yes High With the HPCS With the HPCS SSCs are modeled Spray (HPCS) pressure system system consistent with the TS System inoperable spray into inoperable , two inoperable, one scope and so can be the RPV ECCS low pressure LPCI/LPCS directly included in the pump subsystems subsystem with ERAT for the RICT with ADS actuation manual program.
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| are adequate depressurization E1 - 29
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| GNRO-2023/00014 Page 30 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.5.1.C Two ECCS injection Three LPCI trains, Yes Low pressure With two LPCI/LPCS With two LPCI/ SSCs are modeled subsystems one LPCS train, injection/ subsystems inoperable, LPCS sub- consistent with the TS inoperable and one HPCS spray and either HPCS or two systems inoper- scope and so can be OR train high- remaining LPCI/LPCS able , either HPCS directly included in the One ECCS injection pressure subsystems . or one remaining ERAT for the RICT and one ECCS spray into the LPCI/ LPCS program.
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| spray subsystem RPV PR subsystem inoperable With one HPCS and one OR LPCI subsystem inoperable, two With one HPCS remaining LPCI/LPCS and one LPCI Sub-systems with ADS subsystem actuation (if required). inoperable, one remaining LPCI/LPCS sub-system with manual depress-urization (if required)
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| E1 - 30
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| GNRO-2023/00014 Page 31 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3 .5.1.E One ADS valve Eight ADS Safety Yes Vessel Seven ADS SRVs Automatic and The success criterion is inoperable Relief Valves depressuri- Manual based on plant specific (SRVs) and zation De12ress: MAAP best estimate supporting Four of 8 ADS analysis.
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| components SRVs ELAP depressurization is Manual Deoress very late when decay ELAP: heat is low.
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| One of twenty SRVs_
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| ATWS Pressure Transient:
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| Fifteen of twenty SRVs 3 .5.1.F One ADS valve Eight ADS SRVs, Yes Vessel With one ADS SRV With one ADS SSCs are modeled inoperable . supporting depressuriza- and one low pressure SRV and one consistent with the TS AND components, three tion and low injection/spray LPCI/LPCS scope and so can be One low pressure LPCI trains, and pressure subsystem inoperable, subsystem directly included in the ECCS one LPCS train injection/ the remaining seven inoperable, either ERAT for the RICT injection/spray spray into the ADS SRVs and two of HPCS or one program .
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| subsystem Reactor the remaining low remaining LPCI/
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| inoperable Pressure pressure pump LPCS subsystem Vessel (RPV) systems are adequate with manual depressurization using 4 of the remaining 19 ADS or non-ADS SRVs E1 - 31
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| GNRO-2023/00014 Page 32 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.5.3.A Reactor Core One RCIC train Yes Supply high With the RCIC With the RCIC SSCs are modeled Isolation Cooling pressure system system inoperable, consistent with the TS (RCIC) System makeup inoperable, eith er either HPCS or scope and so can be inoperable water to the HPCS or two one LPCI/LPCS directly included in the RPV. LPCI/LPCS sub-systems with ERAT for the RICT subsystems with manual depress- program.
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| ADS actuation are urization using 4 o adequate. the remaining 19 ADS or non-ADS SRVs 3 .6 .1.2 .C One or more Two double door Isolation To limit One primary None for TS The airlocks are not primary primary function fission containment air lock containment modeled for isolation so a containment air containment air not product door maintains isolation function large pre-existing locks inoperable for locks modeled release primary containment containment isolation reasons other than during and boundary isolation failure will be used as a Condition A or B following on two air locks conservative surrogate postulated for the RICT calculation.
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| Design Basis (See Note 3)
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| Accident (DBAs).
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| E1 - 32
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| GNRO-2023/00014 Page 33 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3 .6 .1.3.A One or more Primary Partial To limit One Primary Same as design Not all primary penetration flow Containment fission Containment success criteria containment isolation paths with one Isolation Valves product Isolation Valve valves are modeled. For PCIV inoperable (PCIVs) release closed per modeled isolation except due to during and penetration pathways, the explicitly leakage not within following modeled valves wi ll be limit. postulated used. For valves th at DBAs are not modeled, a surrogate of a pre-existing containme nt failure is chosen. See Note 4 for details.
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| 3 .6 .1.3.D One or more Primary Yes To limit One Primary Same as design SSCs are modeled penetration flow Containment Purge fission Containment Purge success criteria consistent with the TS paths with one or Valves product Valve closed per scope and so can be more primary release penetration directly included in the containment purge during and ERAT for the RICT valves not within following program.
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| purge valve leakage postulated limits DBAs 3 .6 .1.6 .A One Low-Low Set LLS function of six No LLS function Containment design None Affected SRV will be (LLS) valve safety/relief valves prevents basis of one SRV failed and SRV failure inoperable excessive operating on to reclose probability short duration subsequent actuations will be increased as a SRV cycles is met if five SRVs conservative surrogate with valve actuate at LLS relief (See Note 8) actuation at setpoint.
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| the relief setpoint.
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| E1 - 33
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| GNRO-2023/00014 Page 34 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3 .6 .1.7.A One RHR Two RHR Yes Mitigate the One RHR containment Same as SSCs are modeled containment spray containment spray effects of spray subsystem design success consistent with the TS subsystem subsystems bypass criteria scope and so can be inoperable leakage and directly included in the low energy ERAT for the RICT line breaks program.
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| on containment 3 .6 .2.3.A One RHR Two RHR Yes Removal of One RHR Suppression Same as design SSCs are modeled suppression pool suppression pool heat from the Pool Cooling success criteria consistent with the TS cooling subsystem cooling systems Suppression subsystem scope and so can be inoperable Pool directly included in the ERAT for the RICT program.
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| 3 .6 .2.4 .C One Suppression Two SPMU Yes Provide One SPMU subsystem Same as design SSCs are modeled Pool Makeup subsystems makeup success criteria consistent with the TS (SPMU) subsystem water to the scope and so can be inoperable for Suppression directly included in the reasons other than Pool ERAT for the RICT Condition A or B. program.
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| 3 .6 .5.2 .C Drywell air lock One double door No Isolate drywell One Drywell air lock Same as design The drywell air lock is not inoperable for Drywell air lock boundary to door maintains the success criteria modeled in detail.
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| reasons other than maintain Drywell pressure but screened Therefore, a surrogate is Condition A or B pressure boundary during from modeling. chosen that represents a suppression postulated DBAs failure of containment.
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| function (See Note 5) following postulated DBAs E1 - 34
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| GNRO-2023/00014 Page 35 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3 .6 .5.3.A One or more Drywell isolation Partial Isolate drywell One Drywell Isolation Same as design Not all drywell isolation penetration flow valves boundary to Valve closed per success criteria valves are modeled. For paths with one maintain penetration but screened modeled isolation drywell isolation pressure from modeling. pathways, the explicitly valve inoperable suppression modeled valves will be function used. For unmodeled following valves, the pre-existing postulated large containment DBAs isolation failure (See Note
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| : 6) will be used .
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| 3.7.1 Standby Service Water (SSW) and Ultimate Heat Sink (UHS):
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| 3.7 .1.A One UHS cooling Two cooling Yes To provide One UHS cooling tower Same as design SSCs are modeled tower with one towers, four fans, cooling water fans in either division of success criteria consistent with the TS cooling tower fan and two cooling for the RHR the UHS scope and so can be inoperable tower basins system heat directly included in the exchangers, ERAT for the RICT standby DGs, program.
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| 1--1Prc:: nr-3.7.1.D One SSW Two independent Yes One SSW subsystem Same as design SSCs are modeled I UUI II t;UUlt:I :S ,
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| subsystem cooling water supplying cooling loads success criteria consistent with the TS and ECCS inoperable subsystems (A an d scope and so can be component B) including their directly included in the cooling associated pumps , ERAT for the RICT piping , valves, an d program .
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| instrumentation .
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| E1 - 35
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| GNRO-2023/00014 Page 36 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.8.1 Electrical Power Systems: AC Sources - Operating 3.8.1.A One required offsite From the switchyard Yes ~upply offsite With one required With one required SSCs are modeled circuit inoperable hree electrically and lAC power to offsite circuit offsite circuit consistent with the TS for reasons other physically separated ESF loads inoperable, design inoperable, PRA scope and so can be than Condition F circuits providing AC success criterion is met success criterion directly included in the power to each 4 .16 by the two remaining is met by ERAT for the RICT kV ESF bus. An operable offsite remaining program.
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| offsite circuit consists sources OR two of operable offsite of all breakers, three emergency DGs sources OR one rans formers , capable of supplying emergency DG switches , interrupting associated Division 1, supplying Division de~ces, cab~ng, and 2, or 3 ESF loads. 1 or 2 ESF loads controls required to directly or by ransmit power from crosstie from he offsite Division 3.
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| ransmission network o the onsite Class 1E ESF buses .
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| E1 - 36
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| GNRO-2023/00014 Page 37 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.8 .1.B One required DG Three emergency Yes Supply AC With one required With one required SSCs are modeled inoperable for DGs , their support power when DG inoperable , DG inoperable, the consistent with the TS reasons other than systems, one offsite power design success PRA minimum scope and so can be Condition F emergency DG per ~o ESF 4 .16kV criterion is met by success criterion directly included in the Division 1, 2 and 3 bus is lost two of three offsite is met by offsite ERAT for the RICT ESF load group sources OR two of sources OR one program.
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| three emergency remaining DGs capable of operable supplying associated emergency DG Division 1, 2, or 3 supplying Division ESF loads. 1 or 2 ESF loads directly or by crosstie from Division 3.
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| 3.8.1.C Two required offsite From the switchyard Yes Supply offsite With two required offsite With two required SSCs are modeled circuits inoperable three electrically and AC power to circuit inoperable , design offsite circuits consistent with the TS physically separated ESF loads success criterion is met inoperable, the scope and so can be circuits providing AC by two of three PRAminimum directly included in the power to each 4 .16 emergency DGs capable success criterion ERAT for the RICT kV ESF bus . An of supplying associated is met by the progra m offsite circuit consists Division 1, 2, or 3 ESF remaining offsite of all breakers, loads. source OR one transformers , emergency DG switches , interrupting supplying Division devices, cabling , and 1 or 2 ESF loads controls required to directly or by transmit power from crosstie from the offsite Division 3.
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| transmission network to the onsite Class 1E ESF buses.
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| E1 - 37
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| GNRO-2023/00014 Page 38 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.8.1.D One required offsite From the switchyard Yes Supply offsite With one required offsite With one required SSCs are modeled circuit inoperable three electrically and AC power to circuit and one required offsite circuit and consistent with the TS for reasons other physically separated ESF loads DG inoperable, design one required DG scope and so can be than Condition F circuits providing AC success criterion is met inoperable , PRA directly included in the AND power to each 4 .16 Supply AC by remaining operable success criterion ERAT for the RICT One required DG kV ESF bus . An power when offsite sources OR by is met by program.
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| inoperable for offsite circuit consists offsite power two remaining operable remaining reason other than of all breakers, to ESF emergency DGs capable operable offsite Condition F transformers, 4.16kV bus is of supplying associated sources OR by switches , interrupting lost Division 1, 2, or 3 ESF one remaining de~ces, cab~ng , and loads. operable controls required to emergency DG transmit power from capable of the offsite supplying transmission network associated to the onsite Class Division 1 or 2 1E ESF buses . ESF loads directly or by crosstie from Three emergency Division 3.
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| DGs , their support systems, one emergency DG per Division 1, 2 and 3 ESF load group E1 - 38
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| GNRO-2023/00014 Page 39 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.8.1.F One required Two automatic load Yes Return certain With one required With one required SSCs are modeled automatic load sequencer (one for required plant automatic load automatic load consistent with the TS sequencer each Division 1 and loads in a sequencer (Division 1 or sequencer scope and so can be inoperable 2) predetermine 2) inoperable , the other (Division 1 or 2) directly included in the d sequence to Division (1 or 2) DG AND inoperable, PRA ERAT for the RICT prevent the Division 3 DG are success criterion program .
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| overloading capable of supplying is met by one the power associated Division 1 or remaining source. 2, and 3 ESF loads. operable emergency DG capable of supplying associated Division 1 or 2 ESF loads directly or by crosstie
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| ~rom Division 3.
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| E1 - 39
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| GNRO-2023/00014 Page 40 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.8.4 Electrical Power Systems: DC Sources - Operating 3.8.4.C Division 1 or 2 DC The Division 1 Yes Provide With Division 1 or 2 DC With Division 1 SSCs are modeled electrical power and 2 125 voe DC power electrical power or 2 DC consistent with the TS subsystem busses, breakers, to Division subsystem inoperable , electrical power scope and so can be inoperable for instrumentation , 1 or 2 the design success subsystem directly included in the reasons other than and supports loads criterion is met by the inoperable, the ERAT for the RICT Condition A during remaining divisions of PRA success program.
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| normal electrical power criterion is met and (Division 1 or 2, AND by one of two abnormal Division 3). Division 1 or 2 operation DC electrical power subsystems.
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| Division 3 DC electrical power only required if on AC crosstie or if HPCS is success path.
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| E1 - 40
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| GNRO-2023/00014 Page 41 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.8.7 Distribution Systems-Operating 3.8 .7.A One or more Division 1 and 2 Yes Provide AC With Division 1 or 2 AC With Division 1 or SSCs are modeled Division 1 or 2 AC AC electrical power to the electrical power 2 AC electrical consistent with the TS electrical power distribution associated distribution subsystems power distribution scope and so can be distribution subsystems ESF loads inoperable , the design subsystems directly included in the subsystem(s) consisting of success criterion is met inoperable , the ERAT for the RICT inoperable 4.16Kv buses, by the remaining PRA success program.
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| 480V load divisions of AC electrical criterion is met by centers, power distribution the operable associated subsystems (Division 1 Division 1 or 2 loads, motor or 2, and Division 3) . electrical power control centers , distribution and transformers subsystem.
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| E1 - 41
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| GNRO-2023/00014 Page 42 of 54 Table E1-1: In-scope TS/LCO Conditions to Corresponding PRA Functions GGNS TS Description SSCs Covered by Modeled Function Design Success PRA Success Comments TS TS LCO Condition in PRA Covered by Criteria Criteria TS LCO Condition 3.8 .7.B One or more Division 1 and 2 Yes Provide DC With Division 1 or 2 DC With Division 1 or SSCs are modeled Division 1 or 2 125V DC bus power to the electrical power 2 DC electrical consistent with the TS DC electrical distribution associated distribution power distribution scope and so can be power subsystems ESF loads subsystem(s) subsystems directly included in the distribution inoperable, the design inoperable, the ERAT for the RICT subsystem(s) success criterion is met PRA success program.
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| inoperable by the remaining criterion is met by divisions of DC electrical he operable power distribution Division 1 or 2 DC subsystems (Division 1 electrical power or 2, and Division 3). distribution subsystems.
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| Division 3 DC electrical power only required if on AC crosstie or if HPCS is success path .
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| E1 - 42
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| GNRO-2023/00014 Page 43 of 54 Table E1-1 Notes:
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| : 1. ATWS-RPT system instrumentation is part of the redundant reactivity control system and has two independent trip systems each composed of two channels of each functional input. Each trip system uses a two-out-of-two logic for each function. Thus, either two Reactor Water Level
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| - Low Low, Level 2 or two Reactor Vessel Steam Dome Pressure - High signals are needed to trip a trip system.
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| : 2. There are two HPCS Reactor Vessel Water Level (RVWL) - High (Level 8) channels that provide input to HPCS discharge valve shutoff signal. Failure or unavailability of this function could result in reactor overfill and damage to the RCIC turbine and main turbine, of which RCIC provides a PRA function. Also, the possibility of main-steam isolation would preclude use of the condenser for heat removal. As a conservative surrogate for the maintenance of any TS 3.3.5.1.C function 3.c. high water level channel, RCIC Level 8 logic will be failed, which is a direct failure of RCIC. This is conservative because HPCS and RCIC automatic isolation is assumed failed regardless of a failure of reactor vessel level control and failure of all channels of the trip function. Further, post trip condenser cooling is assumed failed for all TS 3.3.5.1.C function 3.c entries.
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| : 3. The primary containment air locks are not explicitly modeled in the GGNS PRA for TS function to isolate containment and limit releases. It follows that there is no explicit PRA success criterion. However, the LCO condition will be modeled using a pre-existing large containment isolation failure as a conservative surrogate in the PRA. A pre-existing large containment failure event probability was derived by the Pacific Northwest National Laboratory for the NRC (see EPRI Risk Impact Assessment of Extended Integrate Leak Rate Test Intervals, TR-1018243 Rev. 2-A plus the use of NUREG-1493) (References 4 and 5). GGNS is not an outlier in the use of this generic industry accepted data that addresses the operating experience-based probability of containment release pathways being larger than "small".
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| Because the containment hatch doors have no dependencies for the LCO condition, it is appropriate to increase the failure probability of the surrogate event in the ERAT program (versus setting to logical True) for the RICT calculation. This added probability represents the likelihood of failure of the redundant operable door.
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| : 4. Where PCIV SSCs are modeled consistent with the TS scope, unavailability can be directly included in the ERAT for the RICT program. Unmodeled PCIVs were screened in the PRA from LERF consideration based on PC IVs being less than or equal to two inches or if the PCIV isolates a closed system inside containment. However, a conservative assessment using a surrogate pre-existing large containment isolation failure will be used to address individual unmodeled PCIV unavailability (See basis for event in Note 3). Although very conservative, screened penetrations shall be assessed with this surrogate. Where the redundant, unisolated operable isolation valve(s) is(are) fail safe, the respective failure probability(ies) shall be added to the surrogate event in the ERAT program. If any remaining unmodeled, unisolated operable isolation valve is not fail safe, the surrogate shall be set to failed (logical True) for the LCO condition. This approach is conservative because unmodeled PCIV SSCs have been determined not to contribute to LERF.
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| E1-43
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| GNRO-2023/00014 Page 44 of 54
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| : 5. The drywell air locks are not explicitly modeled. A pre-existing containment failure event will be used as a surrogate. Failure of the drywell isolation can result in a loss of vapor suppression for LBLOCA resulting in containment failure during a design basis accident. For the airlocks, the pre-existing large isolation failure event will be used as a surrogate as described in Note 3. This is a suitable surrogate as the consequence of containment release is an equivalent outcome. It is conservative to use this as a surrogate, because only large break LOCA is expected to challenge the vapor suppression function, and the surrogate will be applied to all initiators. This approach is considered conservative as unmodeled isolation paths are screened from LERF concerns in the PRA.
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| : 6. The drywell isolation valves are partially modeled. The pre-existing containment failure event will be used as a surrogate for unmodeled valves. There are two types of drywell isolation valve paths. The first connects the drywell atmosphere to the containment atmosphere (e.g.,
| |
| hydrogen recombiner pathways). Failure of one of these isolation paths could cause a loss of vapor suppression. Therefore, they are treated the same as airlocks in Note 5. The second type of drywell isolation valve connects the drywell through the primary containment to outside the containment. Failure of multiple valves in a path of this type has the consequence of direct release outside containment for open systems. Closed systems will be conservatively treated the same. The applied approach will be the same as described in Note 4. This approach is considered conservative as the unmodeled isolation paths are screened from LERF concerns in the PRA.
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| : 7. Except for MSIVs, one isolation system is associated with the inner primary containment isolation valves and the other isolation system is associated with the outer primary containment isolation valves with the success criteria being closure of one of the two isolation valves. Where SSCs are modeled consistent with the TS scope, SSC will be directly used for unavailability in the ERAT for the RICT program. Otherwise, the use of surrogates will be as follows:
| |
| : a. For functions 1.b through 1.e, MS isolation instrumentation is not explicitly modeled.
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| Function 1.a low reactor water level signal unavailability is modeled under the MSIV logic. This signal function will be used as a surrogate for all MSIV instrumentation functions for this TS (1.b through 1.e). This is a conservative surrogate because, for any function channel maintenance, the entire subsystem is failed and gives no credit for diverse functions in the RICT calculation. The use of the operator action as a surrogate for any of the MSL manual initiating channels is bounding and conservative.
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| : b. The modeled train specific isolation valves associated with the instrumentation subsystem affected by the LCO will be used as a conservative surrogate. This is conservative because it takes no credit for isolation signal diversity or manual action.
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| : c. For unmodeled functions, a large pre-existing containment isolation failure will be used as a conservative surrogate for the LCO condition. See Notes 3 and 4 for the background of the large pre-existing containment failure event and how it will be applied. This approach is conservative because unmodeled isolation functions (i.e.,
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| penetrations/pathways) have been determined not to contribute to LERF.
| |
| : 8. LCOs impacting LLS function of SRV will be treated as the affected SRV is inoperable for all functions AND the SRV Fail to Reclose failure mode probability will be doubled to account for E1-44
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| GNRO-2023/00014 Page 45 of 54 additional SRV cycles. This is conservative, because one LLS valve can be inoperable without impact.
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| : 9. For functions addressing a single channel per ECCS subsystem (train), the success criterion is written in terms of the specific instrument channel and subsystem. With the single channel inoperable, a loss of function as defined by T.S. 3.0.3 does not exist given redundancy and diversity afforded by RCIC, HPCS, LPCS, LPCI A/B/C, ADS and their related instrumentation.
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| : 10. RPS Modeling: The GGNS PRA MOR Rev. 4b models five RPS signals (TS functions) under an AND gate for each channel (A, B, C, and D). These signals include Turbine control/stop, High reactor vessel pressure, low RVWL, High RVWL, and SDV level. It was determined that crediting these five RPS instrumentation signals for any modeled initiating event is non-conservative because not all signals will be challenged for any given initiator. Therefore, these signals will be regrouped according to accident types and conditioned by appropriate initiating events that challenge the signals in the ERAT model.
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| E1-45
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| GNRO-2023/00014 Page 46 of 54 RICTs were calculated for both trains when applicable and the most limiting RICT is specified in Table E1-2. Following implementation of the RICT Program , the actual RICT values will be calculated using the actual plant configuration and the current revision of the PRA model representing the as-built, as-operated condition of the plant, as required by NEI 06-09-A, Revision 0 and the NRC Final Safety Evaluation .
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| RICTs are based on the internal events (including internal flooding) , and internal fire PRA model calculations for CDF and LERF. RICTs calculated to be greater than 30 days are capped at 30 days based on NEI 06-09-A, Revision 0. RICTs not capped at 30 days are rounded to the nearest tenth of a day.
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| Per NEI 06-09-A, Revision 0, for cases where the total CDF or LERF is greater than 1E-03/yr or 1E-04/yr, respectively, the RICT will not be voluntarily entered .
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| Table E1-2: In-Scope TS/LCO Conditions RICT Estimate Tech Spec LCO Condition RICT Estimate 3.1.7.D One SLC subsystem inoperable for reasons other than 30.0 days Condition A, B, or C.
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| 3.3.1.1.A Reactor Protection System (RPS) instrumentation 30.0 days
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| - One or more required channels inoperable 3.3.1.1.B One or more Functions with one or more required channels 30.0 days inoperable in both trip systems.
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| 3.3.4.1.A End of Cycle Recirculation Pump Trip (EOC-RPT) 30.0 days Instrumentation - One or more required channels inoperable.
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| 3.3.4.2.A Anticipated Transient Without SCRAM Recirculation Pump Trip (ATWS-RPT) Instrumentation - One or more channels 30.0 days inoperable 3.3.5.1.B Emergency Core Cooling System (ECCS) Instrumentation -
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| As required by Required Action A.1 and referenced in Tech 30.0 days Spec Table 3.3.5.1-1 .
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| 3.3.5.1.C ECCS Instrumentation - As required by Required Action A.1 30.0 days and referenced in Tech Spec Table 3.3.5.1-1.
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| 3.3.5.1.D ECCS Instrumentation - As required by Required Action A.1 30.0 days and referenced in Tech Spec Table 3.3.5.1-1 .
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| 3.3.5.1.E ECCS Instrumentation - As required by Required Action A.1 30.0 days and referenced in Tech Spec Table 3.3.5.1-1 .
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| 3.3.5.1.F ECCS Instrumentation - As required by Required Action A.1 30.0 days and referenced in Tech Spec Table 3.3.5.1-1.
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| 3.3.5.1.G ECCS Instrumentation - As required by Required Action A.1 30.0 days and referenced in Tech Spec Table 3.3.5.1-1 .
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| E1 -46
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| GNRO-2023/00014 Page 47 of 54 Table E1-2: In-Scope TS/LCO Conditions RICT Estimate Tech Spec LCO Condition RICT Estimate 3.3.5.1.G ECCS Instrumentation - As required by Required Action A.1 30.0 days and referenced in Tech Spec Table 3.3.5.1-1.
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| 3.3.5.3.B Reactor Core Isolation Cooling (RCIC) System Instrumentation
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| - As required by Required Action A.1 and referenced in Tech 30.0 days Spec Table 3.3.5.3-1 3.3.5.3.D RCIC System Instrumentation - As required by Required 30.0 days Action A.1 and referenced in Tech Spec Table 3.3.5.3-1 3.3.6.1.A Primary Containment and Drywell Isolation Instrumentation -
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| 30.0 days one or more required channels inoperable 3.3.6.3.B Residual Heat Removal (RHR) Containment Spray System Instrumentation -As Required by Required Action A.1 and 30.0 days referenced in Tech Spec Table 3.3.6.3-1 3.3.6.3.C Residual Heat Removal (RHR) Containment Spray System Instrumentation -As Requ ired by Required Action A.1 and 30.0 days referenced in Tech Spec Table 3.3.6.3-1 3.3.6.5.A Rel ief and Low-Low Set (LLS) Instrumentation - one trip 30.0 days system inoperable 3.3.8.1.A Loss of Power (LOP) instrumentation - One or more channels inoperable 30.0 days (Referenced in Tech Spec Table 3.3.8.1-1) 3.5.1.A One low pressure ECCS injection/spray subsystem inoperable 30.0 days 3.5.1.B High Pressure Core Spray (HPCS) System inoperable 30.0 days 3.5.1.C Two ECCS injection subsystems inoperable OR 11.4 days One ECCS injection and one ECCS spray subsystem (Note 1) inoperable 3.5.1.E One ADS valve inoperable 30.0 days 3.5.1.F One ADS valve inoperable.
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| AND 30.0 days One low pressure ECCS injection/spray subsystem inoperable 3.5.3.A Reactor Core Isolation Cooling (RCIC) System inoperable 30.0 days 3.6.1.2.C One or more primary containment air locks inoperable for 30.0 days reasons other than Condition A or B 3.6.1.3.A One or more penetration flow paths with one PCIV inoperable 30.0 days except due to leakage not within limit.
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| E1 -47
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| GNRO-2023/00014 Page 48 of 54 Table E1-2: In-Scope TS/LCO Conditions RICT Estimate Tech Spec LCO Condition RICT Estimate 3.6.1.3.A One or more penetration flow paths with one PCIV inoperable 30.0 days except due to leakage not within limit.
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| 3.6.1.3.D One or more penetration flow paths with one or more primary 30.0 days containment purge valves not within purge valve leakage limits 3.6.1.6.A One Low-Low Set (LLS) valve inoperable 30.0 days 3.6.1.7.A One RHR containment spray subsystem inoperable 30.0 days 3.6.2.3.A One RHR suppression pool cooling subsystem inoperable 30.0 days 3.6.2.4.C One Suppression Pool Makeup (SPMU) subsystem inoperable 30.0 days for reasons other than Condition A or B.
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| 3.6.5.2.C Drywell air lock inoperable for reasons other than Condition A 30.0 days or B 3.6.5.3.A One or more penetration flow paths with one drywell isolation 30.0 days valve inoperable 3.7.1.A One UHS cooling tower with one cooling tower fan inoperable 30.0 days 3.7.1.D One SSW subsystem inoperable 16.5 days (Note 2) 3.8.1.A One required offsite circuit inoperable for reasons other than 30.0 days Condition F 3.8.1.B One required DG inoperable for reasons other than Condition 18.7 days F (Note 3) 3.8.1.C Two required offsite circuits inoperable 21 .2 days 3.8.1.D One required offsite circuit inoperable for reasons other than Condition F 13.1 days AND (Note 4)
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| One required DG inoperable for reason other than Condition F 3.8.1.F One required automatic load sequencer inoperable 30.0 days 3.8.4.C Division 1 or 2 DC electrical power subsystem inoperable for 1.6 days reasons other than Condition A (Note 5) 3.8.7.A One or more Division 1 or 2 AC electrical power distribution 2.7 days subsystem(s) inoperable (Note 6) 3.8.7.B One or more Division 1 or 2 DC electrical power distribution 1.2 days subsystem(s) inoperable (Notes 5, 7)
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| E1 -48
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| GNRO-2023/00014 Page 49 of 54 Table E1-2 Notes:
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| : 1. The two limiting configurations are: HPCS and LPCI A out of service (OOS), and LPCI pumps A and B OOS.
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| : 2. SSW B is limiting. SSW A is 30 days.
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| : 3. DG12 is limiting. Other EDGs are 30 days.
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| : 4. DG12 and offsite circuit is limiting. Other configuration RICTs are longer.
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| : 5. This would not be allowed by RICT program, CDF > 1.0E-3.
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| : 6. Most limiting case with AC bus down; this would not be allowed by RICT, CDF > 1.0E-3. Other configurations have RICTs around 30 days.
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| : 7. These RICT cases included two boards/busses; this would not be allowed by RICT, CDF > 1.0E-3. Single panels OOS have longer RICTs.
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| E1 -49
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| GNRO-2023/00014 Page 50 of 54 2.0 ADDITIONAL JUSTIFICATION FOR SPECIFIC ACTIONS This section contains the additional technical justification for the list of Required Actions from Table 1, "Conditions Requiring Additional Technical Justification", of TSTF-505, Revision 2.
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| Additional justification for each of the identified GGNS TS is provided below:
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| Table E1-3: TSTF-505 Rev 2 Table 1 Technical Specifications (TS) that Require Additional Justification TS Description TSTF-505 GGNS TS Additional Justification TS Source Range Monitor 3.3.1.2.A 3.3.1.2.A N/A - TSTF-505 changes are Instrumentation - One excluded.
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| or more required SRMs inoperable in MODE 2 with intermediate range monitors (IRMs) on Range 2 or below.
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| End of Cycle 3.3.4.1.A 3.3.4.1.A TSTF-505 changes are incorporated .
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| Recirculation Pump Trip A loss of trip capability for one or (EOC-RPT) more functions would result in Instrumentation - One or entering Condition B, which would more required channels restore the trip capability or apply inoperable. operational limitations within two hours. A risk informed completion time (RICT) is not being applied to Condition B.
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| EOC-RPT instrumentation is explicitly modeled in the PRA, and all functions of the system are modeled.
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| E1 - 50
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| GNRO-2023/00014 Page 51 of 54 Table E1-3: TSTF-505 Rev 2 Table 1 Technical Specifications (TS) that Require Additional Justification TS Description TSTF-505 GGNS TS Additional Justification TS Relief and Low-Low-Set 3.3.6.5.A 3.3.6.5.A Neither Relief nor LLS (LLS) Instrumentation - instrumentation are modeled One trip system explicitly in the PRA.
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| inoperable For the relief function , division specific SRV pilot SOVs and supports will be added to the SRV relief model as conservative surrogates for the LCO condition. Th is surrogate will have an equivalent impact as channel unavailability.
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| For the LLS functions , the SRV Fail to Reclose failure mode probability will be doubled to account for additional SRV cycles . This is conservative because additional cycles would not occur unless both trip systems were inoperable.
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| The Loss of Power (LOP) 3.3.8.1.A 3.3.8.1.A TSTF-505 changes are incorporated .
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| Instrumentation for each Function in Table 3.3.8.1- A note to preclude RICT entry during 1 shall be OPERABLE - a loss of trip capability is not One or more channels necessary for LCO 3.3.8.1.A as this inoperable. would be addressed by Cond ition B, which requ ires the restoration of actuation capability within one hour; otherwise, the associated diesel generator must be declared inoperable immediately. A RICT is not applied to the Condition B completion time.
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| E1 - 51
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| GNRO-2023/00014 Page 52 of 54 Table E1-3: TSTF-505 Rev 2 Table 1 Technical Specifications (TS) that Require Additional Justification TS Description TSTF-505 GGNS TS Additional Justification TS Primary Containment Air 3.6.1.2.C 3.6.1.2.C TSTF-505 changes are Locks - One or more incorporated. In this Condition, with primary containment air one or more primary containment air locks inoperable for locks inoperable, LCO 3.6.1.2 reasons other than Condition C.1 initiates action to Condition A or B. evaluate the primary containment overall leakage rate per LCO 3.6.1.1.
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| If the overall leakage rate exceeds limits, LCO 3.6.1.1 Condition A would require restoration of primary containment to OPERABLE status within one hour; otherwise, a plant shutdown would be required . A RICT is not proposed for LCO 3. 6.1.1 Completion Times.
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| Primary Containment 3.6.1.3.E 3.6.1.3.D TSTF-505 changes are Isolation Valves (PCIVs) incorporated. In this Condition, with
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| - One or more one or more containment purge penetration flow paths valves not within leakage limits, the with one or more requirements of LCO 3.6.1.1 containment purge valves remain applicable. If containment not within purge valve purge valve leakage is excessive leakage limits. and results in the overall primary containment leakage rate exceed ing limits, then LCO 3.6.1.1 Condition A would require restoration of primary containment to OPERABLE status within one hour; otherwise, a plant shutdown would be required . A RICT is not proposed for LCO 3.6.1.1 Completion Times.
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| Containment Purge Valves are explicitly modeled in the PRA, and all functions of the system are modeled .
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| E1 - 52
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| GNRO-2023/00014 Page 53 of 54 Table E1-3: TSTF-505 Rev 2 Table 1 Technical Specifications (TS) that Require Additional Justification TS Description TSTF-505 GGNS TS Additional Justification TS RHR Containment Spray 3.6.1.7.A 3.6.1.7.A TSTF-505 changes are System - One RHR incorporated.
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| containment spray subsystem inoperable RHR containment spray system is expl icitly modeled in the PRA, and all functions of the system are modeled .
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| Drywell Air Lock System 3.6.5.2.C 3.6.5.2.C TSTF-505 changes are
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| - Drywell air lock incorporated. In this Condition, with inoperable for reasons the drywell air lock inoperable for other than Condition A or reasons other than Condition A or B B, the requirements of LCO 3.6.5.1 (Drywell) remain applicable. If drywell leakage limits are exceeded, then LCO 3.6.5.1 Condition A would require restoration of the drywell to OPERABLE status within one hour; otherwise, a plant shutdown would be required. A RICT is not proposed for LCO 3.6.5.1 Completion Times.
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| Main Turbine Bypass 3.7.6.A 3.7.7.A N/A - TSTF-505 changes are System - Main Turbine excluded.
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| Bypass System inoperable E1 - 53
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| GNRO-2023/00014 Page 54 of 54
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| ==3.0 REFERENCES==
| |
| : 1. Letter from the U.S. NRC to 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' (TAC No. MD4995)", dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| : 2. NEI Topical Report NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines", Revision 0, dated October 2012 (ADAMS Accession No. ML12286A322)
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| : 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. EPRI TR-1018243. Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals. Revision 2-A: June 2008.
| |
| : 5. NUREG-1493. Performance-Based Containment Leak-Test Program. September 1995.
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| E1 - 54
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| Enclosure 2 Grand Gulf Nuclear Station License Amendment Request 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 (3 Pages to Follow)
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| GNRO-2023/00014 Page 1 of 3 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2 1.0 Introduction The NRC Final Safety Evaluation (Reference 1) for NEI 06-09-A (Reference 2), in Item 3 of Section 4, states that the license amendment request (LAR) will provide a discussion of the results of peer reviews and self-assessments conducted for the plant-specific PRA models which support the Risk-Informed Completion Time (RICT) program, including the resolution or disposition of any identified deficiencies.
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| The purpose of this enclosure is to document the technical adequacy of Grand Gulf Nuclear Station (GGNS) PRA models in support of the LAR to adopt TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Revision 2 (Reference 3). Specifically, this enclosure provides a discussion of the results of the peer reviews and self-assessments of the PRA models supporting this application. Note that this enclosure does not provide the resolution or disposition of identified deficiencies as discussed in the NRC Final Safety Evaluation for NEI 06-09-A as all finding-level Facts and Observations (F&Os) were verified to be closed by either a full or focused scope peer review or during a closure review conducted by an independent assessment team. The F&O closure reviews were performed in accordance with Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13 (Reference 4). The NRC found the guidance in Appendix X to be acceptable in its {{letter dated|date=May 3, 2017|text=May 3, 2017 letter}} (Reference 5). The NRCs acceptance acknowledged that F&Os closed in accordance with Appendix X guidance need not be provided in submissions of risk-informed licensing applications.
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| 2.0 PRA Quality/Technical Adequacy The internal events and internal flooding PRA model used to support the RICT program has been assessed against Regulatory Guide (RG) 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2 (Reference 6). The GGNS fire PRA has been assessed against RG 1.200, Revision 3 (Reference 7).
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| The GGNS PRA models are sufficiently robust and suitable for use in risk-informed processes such as regulatory decision-making. The peer reviews that have been conducted and the resolution of findings from those reviews demonstrate that the internal events, internal flooding, and fire PRA models have been performed in a technically sound manner. The assumptions and approximations used in development of the PRA have also been reviewed and are appropriate for their application. Entergy procedures are in place for controlling and updating the models, when appropriate, and for assuring that the model represents the as-built, as-operated plant.
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| The GGNS PRA models of record are maintained as controlled documents and are updated on a periodic basis to represent the as-built, as-operated plant (Reference 8). Entergy procedures provide the guidance, requirements, and processes for the maintenance, update, and upgrade of the PRA.
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| E2 - 1 I
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| GNRO-2023/00014 Page 2 of 3 Internal Events and Internal Flooding PRA Model The GGNS internal events and internal flooding PRA model was subject to a full-scope industry peer review conducted by the Boiling Water Reactor Owners' Group (BWROG) in September 2015. The peer review concluded that the GGNS PRA substantially (approximately 85% of the Supporting Requirements) met the ASME/ANS PRA standard RA-Sa-2009 (Reference 9) at Capability Category II or better.
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| The GGNS PRA internal events model Revision 4a was approved in October 2017 and incorporated changes, as applicable, to support the resolutions of the 2015 peer review findings.
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| The full-scope peer review findings from 2015 were closed by an independent assessment conducted in August 2017. The closure assessment was conducted in accordance with Appendix X to NEI 05-04, 07-12, and 12-13 (Reference 4) utilizing the conditions of acceptance stated in an NRC letter to the Nuclear Energy Institute dated May 2017 (Reference 5). The independent assessment is documented in the closure report and concluded that none of the changes made to the GGNS model were considered a PRA upgrade or use of a new PRA method.
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| Fire PRA Model The fire PRA model was subject to a self-assessment and a full-scope peer review conducted in July 2022 in accordance with NEI 17-07 (Reference 27), ASME/ANS PRA Standard RA-Sa-2009, and Regulatory Guide 1.200, Revision 3. The review was conducted against all technical elements in Part 4 of the ASME/ANS PRA Standard.
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| A finding closure review was conducted on the GGNS fire PRA model in April 2023. The purpose was to perform an independent assessment in accordance with Appendix E of NEI 17-07. The finding-level facts and observations (F&Os) dispositions were reviewed by the independent assessment team; the team determined that all F&Os were adequately addressed and were considered closed. The independent assessment team concurred with the GGNS self-assessment, which concluded that the dispositions for the finding-level F&Os were PRA maintenance activities, and none constituted a PRA model upgrade.
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| Summary The self-assessments and peer reviews demonstrate that the internal events, internal flooding, and fire PRA models are of sufficient quality and level of detail to support the RICT program and have been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC.
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| There are no open finding-level F&Os from the GGNS peer reviews of the PRA models. The results of the reviews have been documented and are available for NRC audit.
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| E2 - 2
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| GNRO-2023/00014 Page 3 of 3 3.0 References
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| : 1. Letter from the NRC to NEI, Final Safety Evaluation for Nuclear Energy Institute (NEI)
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| Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (TAC No. MD4995), dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| : 2. NEI Topical Report NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, dated October 2012 (ADAMS Accession No. ML12286A322)
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| : 3. Letter from the Technical Specification Task Force (TSTF) to the NRC, TSTF Comments on Draft Safety Evaluation for Traveler TSTF-505, Provide Risk-Informed Extended Completion Times and Submittal of TSTF-505, Revision 2, dated July 2, 2018 (ADAMS Accession No. ML18183A493)
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| : 4. NEI letter to NRC, Final Revision of Appendix X to NEI 05-04/07-12/12-16 [sic],
| |
| Close-Out of Facts and Observations (F&Os), dated February 21, 2017 (ML17086A431)
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| : 5. NRC letter to NEI, U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os), dated May 3, 2017 (ML17079A427)
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| : 6. NRC 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)
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| : 7. NRC Regulatory Guide 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 3, dated December 2020 (ADAMS Accession No. ML20238B871)
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| : 8. Entergy Nuclear Management Manual EN-DC-151, PRA Maintenance and Update, Revision 9
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| : 9. The American Society of Mechanical Engineers, 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 2009
| |
| : 10. Nuclear Energy Institute, NEI 17-07, "Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard," Revision 2, dated August 2019 (ADAMS Accession No. ML19241A615).
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| E2 - 3 I
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| Enclosure 3 Grand Gulf Nuclear Station License Amendment Request 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 (1 Page to Follow)
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| GNRO-2023/00014 Page 1 of 1 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 Grand Gulf Nuclear Station (GGNS) submittal. The PRA models used to support the GGNS Risk-Informed Completion Time Program were developed in accordance with PRA standards endorsed in Regulatory Guide 1.200, Revision 2.
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| E3 - 1
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| Enclosure 4 Grand Gulf Nuclear Station License Amendment Request 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 (22 Pages to Follow)
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| GNRO-2023/00014 Page 1 of 22 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
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| : 1. Introduction and Scope Nuclear Energy Institute (NEI) Topical Report NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"
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| Revision 0-A (Reference 1), as clarified by the NRC Final Safety Evaluation (Reference 2),
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| 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, and to discuss conservative 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, and by providing the Grand Gulf Nuclear Station (GGNS)-specific results of the application of the generic methodology and the disposition of impacts on the GGNS Risk-Informed Completion Time (RICT) program. Section 3 of this enclosure presents the plant-specific conservative analysis of seismic risk to GGNS. Section 4 presents the justification for excluding analysis of other external hazards from the GGNS PRA.
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| NEI 06-09-A does not provide a specific list of hazards to be considered in a RICT program.
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| However, non-mandatory Appendix 6-A of the American Society of Mechanical Engineers (ASME) / American Nuclear Society (ANS) RA-Sa-2009 PRA Standard (hereafter "ASME/ANS PRA Standard"), "Addenda to ASME/ANS RA-S-2008 for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear power Plant Applications" (Reference 3) provides a guide for identification of most of the possible external events for a plant site.
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| Additionally, NUREG-1855 Revision 1 (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. These hazards were reviewed for GGNS, along with a review of information pertaining to the site region and plant design to identify the set of external events to be considered. Information from the GGNS Final Safety Analysis Report (FSAR) pertaining to the geologic, seismologic, hydrologic, and meteorological characteristics of the site region, and the current and projected industrial activities in the plant vicinity was reviewed. No new site-specific or plant-unique external hazards were identified through this review. The hazards from Appendix 6-A of the ASME/ANS PRA Standard that were considered for GGNS are summarized in Table E4-1.
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| The scope of this enclosure is consideration of the hazards listed in Table E4-1 for applicability to GGNS. Seismic events are evaluated quantitatively in Section 3, and the other listed external hazards are evaluated and screened as low risk in Section 4.
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| : 2. Technical Approach The guidance contained in NEI 06-09-A states that all hazards that contribute significantly to incremental risk of a configuration must be quantitatively assessed 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 as part of the RICT Program.
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| E4-1
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| GNRO-2023/00014 Page 2 of 22 Consistent with NUREG-1855, Revision 1, external hazards may be addressed as follows:
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| : 1. Screening the hazard based on a low frequency of occurrence,
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| : 2. Conservatively assess the potential impact and including it in the decision-making, or
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| : 3. Developing a PRA model to be used in the RMAT/RICT calculation.
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| The overall process for addressing external hazards considers two aspects of the external hazard contribution to risk.
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| * The first is the contribution from the occurrence of beyond design basis conditions, e.g.,
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| winds greater than design. These beyond design basis conditions challenge the capability of the systems, structures, and components (SSCs) to maintain functionality and support safe shutdown of the plant.
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| * 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 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 in and of itself presents a risk.
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| 2.1 Hazard Screening The first step in the evaluation of the external hazard is screening based on an estimation of a conservative 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)
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| (Reference 5) which is acknowledged in the NRCs Individual Plant Examination of External Events (IPEEE) procedural guidance (Reference 6) as assuring a conservative CDF of less than 1E-06 per year for each hazard. The conservative 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 conservative conditional core damage probability (CCDP) for those conditions. If the conservative CDF for the hazard can be shown to be less than 1E-06 per year, then beyond design basis challenges from the hazard can be screened and do not need to be assessed quantitatively in the RICT Program. The basis for this is as follows:
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| * The overall calculation of the RICT is limited to an incremental core damage probability (ICDP) of 1E-05.
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| * The maximum time interval allowed for the RICT is 30 days.
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| * If the maximum CDF contribution from a hazard is <1E-06 per year, then the maximum ICDP from the hazard is <1E-07 (1E-06/year
| |
| * 30 days/365 days/year).
| |
| * Thus, the conservative ICDP contribution from the hazard is shown to be less than 1% of the permissible ICDP in the conservative time for the condition. Such a minimal contribution is not significant to the decision in computing a RICT.
| |
| The GGNS hazard screening analysis from the IPEEE has been updated to reflect current site conditions. The results are discussed in Section 4 and show that all events listed in Table E4-1 can be screened for GGNS, except for seismic events.
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| While the direct CDF contribution from beyond design basis hazard conditions can be shown to be insignificant using this approach, some external hazards can cause a plant challenge even E4-2
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| | |
| GNRO-2023/00014 Page 3 of 22 for hazard severities that are less than the design basis limit. These considerations are addressed in Section 4.
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| 2.2 Hazard Analysis for CDF Contribution There are two options in cases where the conservative CDF for the external hazard cannot be shown to be less than 1E-06 per year. 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 conservative CDF contribution from the hazard. The conservative approach used to assess the seismic CDF contribution is described in Section 3.
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| 2.3 Evaluation of Conservative Large Early Release Frequency (LERF) Contribution The RICT Program requires addressing both core damage and large early release risk. When a comprehensive PRA does not exist, the 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 conservative estimate of conditional large early release probability (CLERP) that aligns with the conservative CDF evaluation. The incremental large early release frequency (ILERF) is then computed as:
| |
| ILERFHazard = ICDFHazard
| |
| * CLERPHazard The conservative approach used to assess the seismic LERF contribution is described in Section 3.
| |
| 2.4 Risks from Hazard Challenges Upon estimation of a conservative CDF and LERF, the analysis approach must ensure that the RICT Program calculations reflect the change in CDF and LERF caused by out-of-service equipment. As discussed in Section 3, seismic risk is the only beyond design basis hazard that could not be screened out for GGNS. The approach to be used for the RICT program considers that the change in risk with equipment out-of-service will not be higher than the conservative seismic CDF or LERF.
| |
| The above steps address the direct risks from damage to the facility from external hazards.
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| While the direct CDF contribution from beyond design basis hazard conditions can be shown to be insignificant without a full PRA, there may be risks that 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 design basis levels can cause extended loss of offsite power conditions. Additionally, depending on the site, external floods can challenge the availability of normal plant heat removal mechanisms.
| |
| The approach to be 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 the risk. Section 3 provides the analysis of the beyond design basis seismic hazards for the GGNS site, and Section 4 provides an analysis of the representative external hazards for GGNS.
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| E4-3
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| GNRO-2023/00014 Page 4 of 22
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| : 3. Conservative Seismic Analysis This section presents a conservative analysis of the potential seismic impact for inclusion in the decision-making process, as a seismic PRA (SPRA) is not available for GGNS. The process for analyzing an unscreened external hazard without the use of a full PRA involves the following three steps:
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| : 1. Conservatively Estimate CDF
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| : 2. Evaluate Potential Risk Increases Due to Out-of-Service Equipment
| |
| : 3. Qualitatively Evaluate LERF Contribution 3.1 Conservatively Estimate Seismic CDF A seismic PRA is not developed for GGNS. As stated in the IPEEE (Reference 7), GGNS was classified in NUREG-1407 as a reduced-scope plant based on low seismicity. For the IPEEE seismic evaluation, emphasis was placed on conducting detailed seismic walkdowns utilizing the EPRI Seismic Margins Method (SMM). Success Path Logic Diagrams (SPLD) were developed to identify the systems that must function in order to successfully cool the reactor core following the occurrence of the review level earthquake (RLE).
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| The seismic hazard for the GGNS site was re-evaluated in 2014 and provided to the NRC (Reference 8). The peak horizontal acceleration at the site was not expected to exceed 0.10g.
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| However, for additional conservatism a site safe shutdown earthquake (SSE) of 0.15g was selected. For screening purposes, a Ground Motion Response Spectrum (GMRS) was developed, and a probabilistic seismic hazard analysis was completed using the Central and Eastern United States (CEUS) Seismic Source Characterization for nuclear facilities and the updated Electric Power Research Institute (EPRI) Ground-Motion Model. For the 1 to 10 Hz range in the response spectrum, the SSE exceeds the GMRS, which resulted in no risk evaluation being performed. Also, for the portions of higher frequency (>10 Hz), the SSE exceeds the GMRS. Therefore, a high frequency confirmation and spent fuel pool evaluation were not considered necessary.
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| An alternative approach was taken to conservatively estimate SCDF based on the current GGNS seismic hazard curve provided in Reference 8 and conservatively estimate the seismic capacity of a component whose seismic failure would lead directly to core damage. The approach to estimation of the SCDF uses the plant-level high confidence low probability of failure (HCLPF) seismic capacities and convolves the corresponding failure probabilities as a function of the seismic hazard level with the seismic hazard curves from Reference 8. This is a commonly used approach to conservatively estimate SCDF when a seismic PRA is not available; see Section 10-B.9 of the ASME/ANS PRA Standard. This approach is consistent with approaches that have been used in other regulatory applications.
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| It was concluded that GGNS is more robust than was credited in Generic Issue 199 (GI-199)
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| (Reference 9), and the estimated site median capacity and composite uncertainty used by the NRC resulted in a conservative estimate of SCDF. Since no low-capacity significant contributors were identified, the only readily quantifiable source of conservatism is in the NRC assumption that the plant HCLPF is equal to the SSE. Comparing the peak range of the GMRS (1 to 10 Hz) to the 0.8g screening capacity established by the GGNS IPEEE screening and walkdown effort, yields a nominal HCLPF of 0.36g peak ground acceleration (PGA). Given this substantial E4-4
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| GNRO-2023/00014 Page 5 of 22 seismic margin inherent in the GGNS seismic design relative to a state-of-practice seismic hazard, a plant- level HCLPF of 0.3g PGA anchored to the GMRS is a reasonable lower-bound HCLPF. The Grand Gulf plant level HCLPF value of 0.3g is per Engineering Report GGNS-CS-23-00001, "Grand Gulf Seismic High Confidence Low Probability of Failure (HCLPF)
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| Determination," Revision 000, dated April 2023.
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| Using the GGNS HCLPF of 0.3g, the calculations were performed for the PGA hazard, as well as the 25 Hz, 10 Hz, 5 Hz, 2.5 Hz, 1 Hz, and 0.5 Hz spectral frequency hazard curves from Reference 8. Based on these calculations, the 25 Hz seismic hazard curve produces the highest SCDF and is controlling. The total GGNS SCDF is 9.31E-07 per year based on the 25 Hz seismic hazard curve. This SCDF value will be used as the conservative estimate of instantaneous SCDF (ICDFseismic) for the GGNS TSTF-505 LAR RICT calculations.
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| 3.2 Evaluate Potential Seismic Risk Increase Due to Out-of-Service Equipment The approach taken in the computation of SCDF assumes that the SCDF can be based on the likelihood that a single seismic-induced failure leads to core damage. This approach is conservative and implicitly relies on the assumption that seismic-induced failures of equipment show a high degree of correlation (i.e., if one SSC fails, all similar SSCs will also fail). This assumption is conservative, but direct use of this assumption in evaluating the risk increase from out-of-service equipment could lead to an underestimation of the change in risk. If one were to assume no correlation at all in the seismic failures, then the seismic risk would be lower than the risk predicted by a fully correlated model, but the change in risk using the un-correlated model with a redundant piece of important equipment out-of-service would be equivalent to the level predicted by the correlated model.
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| If the industry accepted approach (Reference 10) of correlation is assumed, the conditional core damage frequency given a seismic event will remain unaltered whether equipment is out-of-service or not. Thus, the risk increase due to out-of-service equipment cannot be greater than the total SCDF estimated by the conservative method used in Reference 8. That is, for the GGNS site, the delta SCDF from equipment out-of-service cannot be greater than 9.31E-07 per year.
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| To summarize the above considerations:
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| * The baseline seismic risk in this approach is assumed to be zero, whereas there will always be some level of baseline seismic risk for a zero-maintenance plant configuration. Therefore, the incremental seismic risk (configuration seismic risk baseline seismic risk) will always be overstated using a seismic penalty based on the total estimated seismic risk.
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| * The limiting HCLPF approach assumes that a failure of a component with seismic capacity at that HCLPF leads directly to core damage (CD). However, even common failure of a given set of components (e.g., all emergency diesel generators (EDGs))
| |
| would not lead directly to CD, especially in light of the post-Fukushima FLEX mitigating strategies now in place. In reality, there are few SSCs whose failure would lead to seismic CD with any significant frequency. Examples could be important structures, or the reactor pressure vessel, or "distributed systems" such as all cable trays or all piping systems.
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| E4-5
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| GNRO-2023/00014 Page 6 of 22
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| * In a seismic PRA, seismic impacts to similar components (e.g., all the EDGs) are typically assumed to be correlated unless there are reasons to justify not correlating.
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| Correlation has the effect of introducing common cause impacts. So, if one train of emergency AC power fails seismically, both trains are modeled as likely to fail given the same seismic event. So, in general, most seismic impacts would effectively be equivalent to TS loss of function.
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| * Given the above, the use of a seismic penalty based on assuming seismic core damage given the plant level HCLPF is appropriate.
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| 3.3 Evaluate Seismic LERF Contribution The seismic large early release frequency (SLERF) was conservatively estimated by including the containment fragility in the convolution calculations. The seismic capability of the containment for GGNS was evaluated in the IPEEE. The IPEEE concluded that GGNS was seismically rugged and all components in the SPLD adequately considered the seismic input.
| |
| Since containment is one of the most rigorously seismically designed and analyzed portions of the plant, containment capability is equivalent or greater than the value assessed for plant capability; therefore, a HCLPF of 0.3g can be used as a conservative value to estimate SLERF.
| |
| The corresponding SLERF was calculated by using the plant-level HCLPF as the containment fragility and convolves the corresponding failure probabilities as a function of the seismic hazard level with the seismic hazard curves from Reference 8. The calculations were performed for the PGA hazard, as well as the 25 Hz, 10 Hz, 5 Hz, 2.5 Hz, 1 Hz, and 0.5 Hz spectral frequency hazard curves. The non-seismic containment failure probability (e.g., due to random causes) is calculated based on the ratio of the internal events LERF to the internal events CDF. Based on the internal events LERF of 7.74E-07/yr and the internal events CDF of 2.58E-06/yr, the LERF to CDF ratio is 3.00E-01. The SLERF is the product of the seismic initiating events frequency and the conditional probability of combined core damage and containment failure for each seismic interval. Based on these calculations, the 25 Hz seismic hazard curve produces the highest SLERF and is controlling. The total GGNS SLERF is 7.16E-07 per year based on the 25 Hz seismic hazard curve. This SLERF value will be used as the conservative estimate of instantaneous SLERF (ILERFseismic) for the TSTF-505 submittal RICT calculations.
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| 3.4 Conclusion The above analysis provides the technical basis for addressing the seismic-induced core damage risk for GGNS by reducing the ICDP/ILERP criteria to account for a conservative estimate of the configuration risks due to seismic events.
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| The RICT and RMAT calculations are based on the discussion provided above. The actual RICT and RMAT calculations performed by the GGNS Configuration Risk Management Tool are based on adding an incremental 9.31E-07 per year seismic CDF contribution and a corresponding 7.16E-07 per year seismic LERF contribution to the configuration-specific delta CDF and delta LERF attributed to internal and fire events contributions. This is accomplished by adding these seismic contributions to the instantaneous CDF/LERF whenever a RICT is in effect. This method ensures that an incremental seismic CDF/LERF equal to the conservative SCDF/SLERF is added to internal and fire events incremental CDF/LERF contribution for every RICT occurrence.
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| E4-6
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| GNRO-2023/00014 Page 7 of 22
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| : 4. Evaluation of External Event Challenges and IPEEE Update Results The primary purpose of this section is to address the incremental risk associated with challenges to the facility that do not exceed the design capacity. This section also provides the results of the hazard screening described earlier. Table E4-1 lists the external hazards considered.
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| 4.1 Hazard Screening Except Seismic Events The GGNS IPEEE provides an assessment of the risk to the GGNS associated with external hazards. Additional analyses have been done since the IPEEE to provide updated risk assessments of various hazards, such as aircraft impacts, industrial facilities and pipelines, and external flooding.
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| Table E4-1 reviews the bases for the evaluation of these hazards, identifies any challenges posed, and identifies any additional treatment of these challenges, if required. Table E4-2 provides the criteria applied in the progressive screening process used in this assessment. The conclusions of the assessment, as documented in Table E4-1, ensure that the hazard either does not present a design-basis challenge to GGNS, or is adequately addressed in the PRA.
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| External hazards other than seismic can be screened for the GGNS site.
| |
| In the application of RICTs, 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 GGNS has been performed accounting for such configuration-specific impacts. The process involves multiple steps.
| |
| As a first step in this screening process, hazards that screen for one or more of the following criteria (as defined in Table E4-2) still screen regardless of the configuration, as these criteria are not dependent on the plant configuration.
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| * 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)
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| * The event cannot occur close enough to the plant to affect it. (Criterion C3)
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| * The event which subsumes the external hazard is still applicable and bounds the hazard for other configurations. (Criterion C4)
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| * The event develops slowly, allowing adequate time to eliminate or mitigate the hazard or its impact on the plant. (Criterion C5)
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| 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.
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| As noted above, the configurations to be evaluated are those involving unavailable SSCs whose LCOs are included in the RICT program.
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| E4-7
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| GNRO-2023/00014 Page 8 of 22 Table E4-1: Evaluation of Risks From External Hazards Screening Result External Hazard Screened?
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| Screening Comment (Y/N)
| |
| Criterion Bounding analysis demonstrates that the frequency of aircraft-induced radiological consequences is less than 1E-7/yr. Conservatively assuming the Large Early Release Frequency (LERF) is a surrogate for the radiological consequences and that Aircraft Impact Y PS4 Core Damage Frequency (CDF) is typically an order of magnitude greater than LERF for Boiling Water Reactors, this implies that CDF is less than 1E-6/yr. Therefore, the aircraft impact hazard can be screened out from an external events PRA for Grand Gulf Nuclear Station (GGNS).
| |
| Topography is such that no avalanche Avalanche Y C3 is possible.
| |
| The hypochlorite system inhibits growth and is controlled and monitored. There would be adequate Biological Event Y C1 warning for these events. Also note that the Mississippi River is not the Ultimate Heat Sink (UHS) for GGNS.
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| The GGNS site is inland just east of Coastal Erosion Y C3 the Mississippi River.
| |
| There is no surface water intake for the plant located in the river which would be affected by low flow in the river. Makeup water is supplied by the C1 radial collector well system. However, Drought Y the radial collector wells are not C5 required to be operable following an accident because the standby service water basins have sufficient capacity for 30 days of operation and makeup is not required during that period.
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| E4-8
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| GNRO-2023/00014 Page 9 of 22 Screening Result External Hazard Screened?
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| Screening Comment (Y/N)
| |
| Criterion The updated examination of external flood risk, including the updated plant data, flood history and new measures for risk management validate the External Flooding and C1 current flood mitigation strategy of the Y
| |
| Intense Precipitation PS1 current design basis. The results of the flooding hazard reevaluation report indicate that external flooding events will cause no flooding damage to GGNS safety-related SSCs.
| |
| All Seismic Category I structures are designed for the 100-year wind speed of 90 mph sustained at 30 ft above grade. Tornado loadings are based on a 290-mph rotational wind speed and a 70-mph translational wind speed, with a simultaneous maximum atmospheric pressure drop of 3 psi at a rate of 2 psi/second. Non-Category I structures have been designed to not collapse on or impact Seismic Category I structures. According to the original analysis performed to verify compliance with the licensing basis, there are no openings in the walls or Extreme Wind or roofs of Seismic Category I structures Y PS4 Tornado which could allow a tornado missile to pass through and hit any safety-related targets.
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| A Tornado Missile Risk Evaluator (TMRE) analysis was performed using NEI 17-02 methodology to determine the missile strike probability for identified openings.
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| The results of the TMRE analysis were that not providing missile protection to preclude damage related to nonconforming SSCs represented a risk increase of 1.35E-07/yr CDF and 1.25E-09/yr LERF, which was sufficiently low to conclude that E4-9
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| GNRO-2023/00014 Page 10 of 22 Screening Result External Hazard Screened?
| |
| Screening Comment (Y/N)
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| Criterion additional tornado missile protection need not be provided.
| |
| In summary, GGNS has been designed for extreme winds and tornado loadings that are higher than the current regulatory guidance. A TMRE analysis was completed for licensing basis deviations, and the total risk increase was very small.
| |
| Therefore, it is concluded that the hazard event of extreme winds and tornadoes can be screened out for GGNS.
| |
| It affects frequency of occurrence of other hazards, e.g., highway Fog Y C4 accidents, aircraft landing and take-off accidents and is indirectly considered.
| |
| Forest fires have been recorded in Claiborne County; however, the size of these fires is relatively small. Most fires were reported to be caused by farmers burning trash but given the marshy nature of the region, fires are not expected to burn beyond control.
| |
| Approximately 200 acres of land around the plant are cleared, thus C1 eliminating potential for a forest fire in Forest or Range Fire Y the immediate proximity of the plant C3 site. If such an event were to occur, self-contained breathing apparatuses are available for emergency needs.
| |
| Also, the main control room (MCR) can be manually isolated as necessary well in advance of the detection of smoke surrounding the facility.
| |
| Therefore, the release of toxic combustion products from forest fire does not pose a hazard to MCR personnel, nor will it cause thermal E4-10
| |
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| GNRO-2023/00014 Page 11 of 22 Screening Result External Hazard Screened?
| |
| Screening Comment (Y/N)
| |
| Criterion damage to the GGNS safety-related structures.
| |
| Frost Y C4 Snow and ice govern this hazard.
| |
| Hail may occur, but there are no openings in the walls or roofs of safety related buildings through which hail may enter and damage essential equipment. Hail occurs at the site about two days per year, with the most probable month of occurrence being April. Hail of damaging nature occurs C1 infrequently. Section 3.5.1 of the Hail Y UFSAR identifies the systems that C4 must be protected from missiles (both internally and externally). Tornado-generated missiles, which are considered as the limiting natural hazard, provide for rigorous design of the protective structures which house critical components. The effects from hail are bounded by the current design basis of the plant.
| |
| The highest recorded air temperature near Port Gibson was 107°F on August 18, 1909. The maximum allowable cold-water temperature is nominally 90°F. The highest recorded water temperature in Port Gibson, MS does not approach maximum High Summer C1 allowable temperature for containment Y heat removal systems given that water Temperature C5 heats and cools slower than air.
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| Thermal inertia of concrete structures where safety-related equipment is located should not have an impact.
| |
| The longest period in Port Gibson with max temperatures of 90°F or higher on successive days was 42 days (June 28, 2022 - August 8, 2022) with a E4-11
| |
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| GNRO-2023/00014 Page 12 of 22 Screening Result External Hazard Screened?
| |
| Screening Comment (Y/N)
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| Criterion maximum daily average of 86.5°F on July 9, 2022.
| |
| GGNS is located about 150 miles High Tide, Lake Level, inland from the Gulf Coast and over a Y C3 or River Stage mile away from the Mississippi River.
| |
| GGNS is not affected by High Tide.
| |
| The plant is about 150 miles inland from the Gulf Coast; however, wind speeds generated during hurricanes and other storms are less intense and lower in magnitude than those generated by tornadoes. Thus, plant structures that are designed to satisfy Hurricane Y C1 the design criteria for tornadoes will also satisfy the design criteria for those events categorized as "high winds". In addition, a hurricane undergoes significant weakening by the time it reaches the GGNS, which is about 150 miles inland.
| |
| Since the plant is subjected to a subtropical climate with mild winters, C1 prolonged snowfalls or large Ice Cover Y accumulations of snow or ice on the C4 ground and structures are not anticipated. No damage from snow or ice loading on structures is expected.
| |
| Military Facility Accident There are no military facilities within five miles of GGNS. Therefore, this potential contributor to risk was eliminated.
| |
| Industrial or Military C3 Y Industrial Facility Accident Facility Accident PS1 According to the IPEEE, there is no extensive industrial activity around the Grand Gulf site. There are no military installations, chemical or munition plants, stone quarries, or major E4-12
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| GNRO-2023/00014 Page 13 of 22 Screening Result External Hazard Screened?
| |
| Screening Comment (Y/N)
| |
| Criterion gasoline-storage areas located within 5 miles of the plant.
| |
| When the IPEEE was completed, the nearest industrial facility that had significant quantities of stored chemicals is in Port Gibson about 4.5 miles away. This facility has been closed and inoperable since June 2002. The IPEEE remains bounding.
| |
| Flammable Stationary Sources The IPEEE states that a 4-inch gas line exists approximately 4.5 miles east of GGNS. No new oil or gas pipelines or associated oil and gas facilities have been constructed near GGNS since the IPEEE was completed. The IPEEE remains bounding.
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| The GGNS internal events PRA model Internal Flooding N None includes evaluation of risk from internal flooding events.
| |
| The GGNS fire PRA model includes Internal Fire N None evaluation of risk from internal fires.
| |
| Terrain surrounding the immediate Landslide Y C3 vicinity of the plant is such that landslide is highly unlikely.
| |
| C1 Lightning is considered in the plant Lightning Y design. It is also considered in the C4 LOOP analysis for the site PRA.
| |
| The UHS, consisting of the standby service water (SSW) system cooling C1 towers and makeup basins, provides Low Lake Level or heat rejection and makeup water Y
| |
| River Stage C5 required for the dissipation of heat to permit safe shutdown and cooldown of the plant and to maintain it in a safe shutdown condition. GGNS does not E4-13
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| GNRO-2023/00014 Page 14 of 22 Screening Result External Hazard Screened?
| |
| Screening Comment (Y/N)
| |
| Criterion rely on the Mississippi River for the UHS.
| |
| The lowest recorded temperature in Port Gibson was -5°F on January 27, 1940. The lowest average monthly temperature was observed to be 33.8°F also in January of 1940. The cooling towers and makeup basins are Low Winter relatively insensitive to extreme Y C1 Temperature temperatures. Failures due to cold are generally limited to full or partial loss of offsite power, which is sufficiently analyzed in the internal events PRA.
| |
| Severe temperature transients are, therefore, eliminated from any further analysis.
| |
| This event has a very low frequency of Meteorite or Satellite occurrence for any site in the United Y C2 Impact States.
| |
| A 4-inch gas line exists approximately 4.5 miles east of GGNS. The pipelines Pipeline Accident Y C3 are located beyond the 2-mile search area criterion.
| |
| Chemicals on the GGNS site were analyzed in accordance with Regulatory Guide 1.78.
| |
| The GGNS Chemical Control Program regulates the use, storage and disposal of chemicals present on site.
| |
| New chemicals are required to be Release of Chemicals evaluated for toxic characteristics Y C4 in Onsite Storage under this program, and to have a control room habitability evaluation performed, if applicable. No new threats that would challenge control room habitability in event of a postulated accident were found.
| |
| Therefore, onsite sources of toxic chemicals do not pose a threat to E4-14
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| GNRO-2023/00014 Page 15 of 22 Screening Result External Hazard Screened?
| |
| Screening Comment (Y/N)
| |
| Criterion control room habitability and the IPEEE remains bounding.
| |
| The Mississippi River is not used as River Diversion Y C3 the UHS; therefore, this event will not impact GGNS.
| |
| This is not a relevant hazard for this Sand or Dust Storm Y C3 region.
| |
| There is no large body of water capable of producing a seiche within Seiche Y C3 50 miles of the site; therefore, a seiche cannot affect the plant.
| |
| Seismic activity is not screened out in this analysis but is addressed on a Seismic Activity N None case-by-case basis depending on the requirements of the specific risk-informed program.
| |
| Snowfall in the plant area occurs about once a year with an average depth of 2 inches. The site is not subject to heavy snow accumulations. In assessing the adequacy of the site and roof drainage systems during winter probable maximum precipitation, they are conservatively assumed to be blocked with ice.
| |
| During this condition the quantity of Snow Y C3 rain that falls on the site, including spillover from building roofs through the scuppers, will flow away from plant buildings as surface runoff.
| |
| For the period between 1894 and 2022, the maximum observed snowfall for any given year occurred in 1967 with a total of 10 inches between February and March. Assuming no snow melt occurred between these two discreet occurrences, the E4-15
| |
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| GNRO-2023/00014 Page 16 of 22 Screening Result External Hazard Screened?
| |
| Screening Comment (Y/N)
| |
| Criterion maximum depth would be 10 inches.
| |
| According to UFSAR Section 2.3.1.2.6, the SR structures at GGNS are designed to withstand 10 inches of snow.
| |
| The 48-hr probable maximum winter precipitation (PMWP) for the site was conservatively estimated to be 7 inches (water equivalent) in UFSAR Section 2.3.1.2.5. Based on the record of historical snow and ice storms which have affected the area, it is unlikely that a significant percentage of the PMWP would fall in a frozen form which would result in a roof load (assuming runoff) exceeding design limit.
| |
| The Catahoula Formation of Miocene age is the foundation-bearing stratum for the major plant structures. It consists primarily of hard-to-very-hard silty-to-sandy clay, clayey silt, and locally indurated or cemented clay, silt, and sand layers. The maximum settlement expected is less than one Soil Shrink-Swell Y C3 inch. Structural backfill is sand Consolidation compacted to a minimum 95 percent ASTM D-1557 dry density. The minimum factor of safety against liquefaction is 1.8. These are strong, statically, and dynamically stable materials. Due to the permeable nature of the granular soils at the site, the soil is resistant to shrink-swell.
| |
| Storm Surge Y C4 Included in external flooding analysis.
| |
| Toxic gas is covered under release of chemicals in onsite storage, industrial Toxic Gas Y C4 or military facility accident, and transportation accident.
| |
| E4-16
| |
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| GNRO-2023/00014 Page 17 of 22 Screening Result External Hazard Screened?
| |
| Screening Comment (Y/N)
| |
| Criterion Shipments by River The nearest route to GGNS by river transport is the Mississippi River. The Mississippi River passes 1.34 miles west of the site at its closest point.
| |
| Furthermore, most of the plant structures and all of the safety-related structures are located on top of a bluff approximately 65 feet above the normal river level. The bluff provides an earthen shield against explosions of potential river-traffic cargo. Also, any flammable vapor clouds released by a traffic accident in the river would be partially shielded by the high elevated riverbank. Therefore, no explosive hazard to the plant located 1.34 miles inland is anticipated due to C3 Transportation an ignition of the vapor cloud. The Y
| |
| Accident PS1 IPEEE remains bounding.
| |
| Shipments by Rail The nearest railroad carrying hazardous material is 30 miles south of the site. Based on the separation distances, it is concluded that accidents along this railroad will not affect the safe operation of the plant.
| |
| The Mississippi Department of Transportation State Rail Plan Update was reviewed, and no new railways were identified near GGNS. The IPEEE remains bounding.
| |
| Shipments by Truck The nearest highway on which explosive materials may be transported is U.S. Highway 61, which is a minimum distance of 4.5 miles from the center of the reactor. Based E4-17
| |
| | |
| GNRO-2023/00014 Page 18 of 22 Screening Result External Hazard Screened?
| |
| Screening Comment (Y/N)
| |
| Criterion on the separation distances, it is concluded that accidents along this highway will not affect the safe operation of the plant. No new mobile source pathways or significant changes in hazardous materials transported by highway within a 5-mile radius of GGNS were identified. The IPEEE remains bounding.
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| Toxic Gas from Transient Sources The IPEEE states that the risk to control room operators by the effects of toxic clouds resulting from barge accidents is acceptably low due to the low concentration of the chemical upon reaching the control room or the low probability of an accident which falls within the guidelines established in the Standard Review Plan section 2.2.3. The most current version of this section has been reviewed. The guidelines are consistent with the previous revision. The nearest major highway is approximately 4.5 miles east-southeast of the site, and the nearest railroad carrying hazardous materials is 30 miles south of the site.
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| This remains consistent with the IPEEE, in which it was concluded that accidents along these routes will not affect the safe operation of the plant based on the separation distances.
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| Therefore, transient sources of toxic chemicals do not pose a threat to control room habitability and the IPEEE remains bounding.
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| GGNS is located about 150 miles Tsunami Y C3 inland. The site is not exposed to the tsunami threat.
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| E4-18
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| GNRO-2023/00014 Page 19 of 22 Screening Result External Hazard Screened?
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| Screening Comment (Y/N)
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| Criterion Analysis determined that for the Turbine-Generated Y C1 current design this is not a credible Missiles hazard at GGNS.
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| The site is not close to any active Volcanic Activity Y C3 volcanoes.
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| Waves Y C4 Included in external flooding analysis.
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| E4-19
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| GNRO-2023/00014 Enclosure 4 Page 20 of 22 Table E4-2: Progressive Screening Approach for Addressing External Hazards Event Analysis Criterion Source Comments C1. Event damage potential is NUREG/CR-2300 and less than events for which plant is ASME/ANS Standard designed. RA-Sa-2009 C2. Event has lower mean NUREG/CR-2300 and frequency and no worse ASME/ANS Standard consequences than other events RA-Sa-2009 analyzed.
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| Initial Preliminary C3. Event cannot occur close NUREG/CR-2300 and Screening enough to the plant to affect it. ASME/ANS Standard RA-Sa-2009 C4. Event is included in the NUREG/CR-2300 and Not used to screen.
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| definition of another event. ASME/ANS Standard Used only to include RA-Sa-2009 within another event.
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| C5. Event develops slowly, ASME/ANS Standard allowing adequate time to RA-Sa-2009 eliminate or mitigate the threat.
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| PS1. Design basis hazard cannot ASME/ANS Standard cause a core damage accident. RA-Sa-2009 PS2. Design basis for the event NUREG-1407 and meets the criteria in the NRC ASME/ANS Standard 1975 Standard Review Plan RA-Sa-2009 Progressive (SRP).
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| Screening PS3. Design basis event mean NUREG-1407 as frequency is < 1E-05 per year modified in and the mean conditional core ASME/ANS Standard damage probability is < 0.1. RA-Sa-2009 PS4. Bounding mean CDF is < NUREG-1407 and 1E-06 per year. ASME/ANS Standard RA-Sa-2009 Screening not successful. PRA NUREG-1407 and Detailed PRA needs to meet requirements in ASME/ANS Standard the ASME/ANS PRA Standard. RA-Sa-2009 E4-20
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| GNRO-2023/00014 Page 21 of 22 4.2 Seismically-Induced Loss of Offsite Power Challenges For the GGNS site, the only incremental risk associated with challenges to the facility that do not exceed the design capacity that is not already addressed is seismically-induced LOOP. The GGNS seismic LOOP CDF results were conservatively estimated by including the diesel generator and offsite power fragility in the convolution calculations. The approach to estimation of the seismic LOOP CDF uses the diesel generator and offsite power HCLPF seismic capacities obtained from the NRC Risk Assessment of Operational Events Handbook (Reference 12) and convolves the corresponding failure probabilities as a function of the seismic hazard level with the GGNS 25 Hz seismic hazard curve.
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| The internal events CDF contribution due to LOOP is 8.42E-07 and the estimated seismic-induced LOOP CDF contribution is 5.49E-08; therefore, the seismic-induced LOOP frequency is less than 7% of the total LOOP frequency that is already accounted for in the internal events PRA. This frequency is judged to be a sufficiently small fraction that it will not significantly impact the RICT program calculations and can be omitted.
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| : 5. Conclusions Based on this analysis of external hazards for GGNS, no additional external hazards need to be added to the existing PRA models. The evaluation concluded that the hazards either do not present a design-basis challenge to GGNS, the challenge is adequately addressed in the PRA, or the hazard has a negligible impact on the calculated RICT and can be excluded.
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| The ICDP/ILERP acceptance criteria of 1E-05/1E-06 will be used within the RICT Program framework to calculate the resulting RICT and RMAT based on the total configuration-specific delta CDF/LERF attributed to internal events and internal fire, plus the conservative seismic delta CDF/LERF values.
| |
| : 6. References
| |
| : 1. Nuclear Energy Institute (NEI) Topical Report (TR) 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 (part of ADAMS Package Accession No. ML122860402))
| |
| : 2. Letter from the NRC to 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 (TAC No. MD4995)," 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
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| : 4. NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," Revision 1, March 2017 E4-21
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| | |
| GNRO-2023/00014 Page 22 of 22
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| : 5. NUREG-75/087, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," 1975
| |
| : 6. NUREG-1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," June 1991
| |
| : 7. Entergy Letter to NRC, "Individual Plant Examination of External Events (IPEEE) Schedule Final Submittal," dated November 15, 1995
| |
| : 8. Entergy Letter to NRC, "Entergy Seismic Hazard and Screening Report (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. ML14090A098)
| |
| : 9. NRC Generic Issue 199 (GI-199) "Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, Safety/Risk Assessment," dated August 2010 (ADAMS Accession No. ML100270639)
| |
| : 10. Letter from the Electric Power Research Institute (EPRI) to the Nuclear Energy Institute (NEI), "Fleet Seismic Core Damage Frequency Estimates for Central and Eastern U.S.
| |
| Nuclear Power Plants Using New Site-Specific Seismic Hazard Estimates," dated March 11, 2014 (ADAMS Accession No. ML14080A589)
| |
| : 11. Nuclear Energy Institute (NEI), NEI 17-02, "Tornado Missile Risk Evaluator (TMRE) Industry Guidance Document," Revision 1, dated September 2017 (ADAMS Accession No. ML17268A036)
| |
| : 12. NRC Handbook, "Risk Assessment of Operational Events Handbook, Volume 2 - External Events, Internal Fires - Internal Flooding - Seismic - Other External Events, Frequencies of Seismically-Induced LOOP Events," Revision 1.02, dated November 2017 (ADAMS Accession No. ML17349A301)
| |
| : 13. NUREG/CR-2300, "PRA Procedures Guide: A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants," January 1983 E4-22
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| | |
| Enclosure 5 Grand Gulf Nuclear Station License Amendment Request Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b Baseline CDF and LERF (2 Pages to Follow)
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| GNRO-2023/00014 Page 1 of 2 Baseline CDF and LERF 1.0 Introduction Section 4.0, Item 6 of the NRC Final Safety Evaluation (Reference 1) for NEI 06-09-A (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 that these values are within the guidelines established in Regulatory Guide (RG) 1.174, Revision 1 (Reference 3). Note that RG 1.174, Revision 2 (Reference 4) and RG 1.174, Revision 3 (Reference 5) did not revise these guidelines.
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| The purpose of this enclosure is to demonstrate that the Grand Gulf Nuclear Station (GGNS) total CDF and total LERF are within the guidelines established in RG 1.174, which does not establish firm limits for total CDF and LERF, but 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 guidelines are met confirms that the risk metrics of NEl-06-09-A can be applied to the Risk-Informed Completion Time (RICT) program.
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| 2.0 Baseline Risk Table E5-1 provides the GGNS CDF and LERF values that resulted from a quantification of the baseline internal events, internal flooding, and fire PRA models. This table also includes a conservative estimate of the seismic contribution to CDF and LERF based on the methodology detailed in Enclosure 4. Other external hazards are below accepted screening criteria and therefore do not contribute significantly to the totals. Note that the values in Table E5-1 were quantified using average maintenance (i.e., nominal maintenance was used, whereas the application specific model will be a zero-maintenance model).
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| Table E5-1 GGNS Total Baseline CDF/LERF Source CDF Contribution LERF Contribution (per year) (per year)
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| Internal Events PRA 2.12E-06 5.11E-07 Internal Flooding PRA 4.65E-07 2.63E-07 Fire PRA 2.22E-05 2.87E-06 Seismic Penalty 9.31E-07 3.58E-07 Total 2.57E-05 4.00E-06 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, GGNS TSTF-505 implementation is consistent with NEI 06-09-A guidance.
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| E5 - 1
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| GNRO-2023/00014 Page 2 of 2 3.0 References
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| : 1. Letter from the NRC to NEI, Final Safety Evaluation for Nuclear Energy Institute (NEI)
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| Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (TAC No. MD4995), dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| : 2. Nuclear Energy Institute (NEI) Topical Report NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, dated October 2012 (ADAMS Accession No. ML12286A322)
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| : 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, November 2002 (ADAMS Accession No. ML023240437)
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| : 4. NRC Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 2, dated May 2011 (ADAMS Accession No. ML100910006)
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| : 5. NRC 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)
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| E5 - 2
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| | |
| Enclosure 6 Grand Gulf Nuclear Station License Amendment Request 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 (1 Page to Follow)
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| | |
| GNRO-2023/00014 Page 1 of 1 Justification of Application of At-Power PRA Models to Shutdown Modes This enclosure is not applicable to the Grand Gulf Nuclear Station submittal. Entergy is proposing to apply the Risk-Informed Completion Time Program only in Modes 1 and 2 and not in the shutdown Modes.
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| E6 - 1
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| Enclosure 7 Grand Gulf Nuclear Station License Amendment Request 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 (4 Pages to Follow)
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| GNRO-2023/00014 Page 1 of 4 PRA Model Update Process 1.0 Introduction Section 4.0, Item 8 of the NRC Final Safety Evaluation (Reference 1) for NEI 06-09-A (Reference 2) requires that the license amendment request (LAR) provide a discussion of the licensees programs and procedures which assure the PRA models that support the Risk-Managed Technical Specifications (RMTS) are maintained consistent with the as-built, as-operated plant.
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| Entergy utilizes a formal process to maintain and update the PRA models such that the models reflect the as-built, as-operated plant. The PRA models of record will be used as the basis for the application specific model that supports the Risk-Informed Completion Time (RICT) program at the Grand Gulf Nuclear Station (GGNS). This enclosure describes the procedural processes that will be used to ensure the configuration control of models used to support the RICT program. Plant changes, including physical modifications and procedure revisions, will be identified and reviewed prior to implementation to determine if the change could impact the PRA models. Entergys PRA configuration control process is governed by procedure EN-DC-151, PRA Maintenance and Update (Reference 3). The process ensures that plant modifications, procedure revisions, and other items are incorporated into the PRA models as appropriate. The process will include monitoring for and evaluation of conditions that may impact the PRA models (including potential or known errors); issues that are determined to be conditions adverse to quality will be entered into the stations Corrective Action Program.
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| If it is identified that a plant change or discovered condition has a significant impact to RICT program calculations as defined by procedure, then an unscheduled, interim update to the PRA model will be implemented. Otherwise, the change will be incorporated during the next periodic model update. Such pending changes are considered when evaluating other changes until they are fully implemented into the PRA models. Periodic updates will typically be performed every two fuel cycles.
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| 2.0 PRA Model Update Process 2.1 Internal Events, Internal Flooding, and Fire PRA Model Maintenance and Update The Entergy fleet PRA model maintenance and update process will ensure that the applicable PRA models of record and application specific model used for the RICT program reflect the as-built, as-operated plant at GGNS. The process delineates the responsibilities and guidelines for updating the full power internal events, internal flooding, and fire PRA models, and includes both periodic and interim updates.
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| The process will include provisions to:
| |
| * track, evaluate, and prioritize issues that may affect 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);
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| E7 - 1
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| GNRO-2023/00014 Page 2 of 4
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| * assess the individual and cumulative risk impact of unincorporated changes; and
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| * control the model and necessary computer files, including those associated with the Configuration Risk Management Program (CRMP) model.
| |
| Industry best practices and consensus modeling techniques are also reviewed and monitored to ensure Entergys PRA is using state of the art processes and methods. Changes that are considered an upgrade per the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by Regulatory Guide 1.200, Revision 3) will receive a peer review focused on those aspects of the PRA model that represent the upgrade.
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| 2.2 Review of Plant Changes for Incorporation into the PRA Model
| |
| : 1. Entergy utilizes a Model Change Request (MCR) database to track all PRA model changes including physical modifications to the facility and changes to operating practices and procedures. Changes with potential significant risk impact are tracked using the MCR database, but they may also be entered into the Corrective Action Program (if the issue rises to the level of a condition adverse to quality).
| |
| : 2. Plant changes or discovered conditions are captured in the MCR database and reviewed to determine the potential impacts to the PRA models as well as the CRMP model and the subsequent risk calculations that support the RICT program (NEI 06-09-A, Section 2.3.4, Items 7.2 and 7.3, and Section 2.3.5, Items 9.2 and 9.3).
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| : 3. Entergy uses a grading system to evaluate and prioritize the issues documented in the MCR database. Any MCR that receives a grade of A requires an interim model update.
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| MCR grade A is defined as Extremely important and necessary to assure the technical adequacy of the PRA or quality of the PRA. This model update would include an update of the impacted applications, the CRMP model used to support the RICT program being one of those applications. Upon determining an MCR database entry is grade A, the issue is entered into the Corrective Action Program by initiating a Condition Report.
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| : 4. Plant changes will be evaluated and screened based on risk criteria consistent with procedural requirements with consideration of the cumulative impact of other pending changes. MCR database entries with the potential for significant impact to the CRMP tool used for performing RICT calculations will be incorporated as an interim, unscheduled update (grade A prioritization). The timeliness of the unscheduled update will be commensurate with the significance of the issue(s).
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| : 5. If planned changes or errors have a significant impact to the PRA model, which will be defined as a greater than 25% increase in core damage frequency (CDF) or large early release frequency (LERF), or significant impacts to basic event importance measures (a factor of three increase in the corrected Birnbaum value of a monitored MSPI train or component) used for RICT calculations, then an unscheduled update will occur to update the model of record and/or application specific model, as appropriate.
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| If the 25% increase in CDF or LERF criteria is exceeded, then use of the RICT program is suspended until the issue can be addressed, except when the deviation is such that impacted RICTs remain conservative. The PRA engineer may also perform and document a standalone, interim analysis that justifies continued use of the RICT E7 - 2
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| GNRO-2023/00014 Page 3 of 4 program if the results of the analysis bound the issue documented in the MCR database.
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| For example, the interim analysis could involve additional PRA refinement to model the system and/or issue in greater detail. This interim analysis is also discussed in item 8 of this section. The station will move forward with an unscheduled PRA update if the 25%
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| CDF or LERF criteria are exceeded regardless of whether interim analyses justify continued use of the RICT program (i.e., interim analyses do not allow deferring an unscheduled PRA update when the criteria are exceeded, but these analyses can defend continued use of the program while the unscheduled PRA update is being implemented).
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| If it is not practical to assess the impact quantitatively, then a qualitative assessment, utilizing the experience and judgment of the PRA engineer, is performed considering the potential change in basic event importance measures for each application. This assessment utilizes the experience and judgment of the PRA engineer to determine if there are any issues that are individually negligible but could collectively impact the RICT program.
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| : 6. Otherwise, the change will be assigned a priority and will be incorporated during a subsequent update consistent with procedural requirements.
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| : 7. PRA model of record updates will typically be performed once every two fuel cycles but may be completed sooner depending on the changes or issues documented in the MCR database and at the discretion of management (NEI 06-09-A, Section 2.3.4, Item 7.1, and Section 2.3.5, Item 9.1).
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| : 8. If a PRA model change is required for the CRMP model, but cannot be immediately implemented for a significant plant change or discovered condition, either one of the following is applied:
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| : a. Analysis 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. The use of such bounding analyses is consistent with the guidance of NEI 06-09-A.
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| or
| |
| : b. Appropriate administrative restrictions on the use of the RICT program for extended Completion Time are put in place until the model changes are completed, consistent with the guidance of NEI 06-09-A.
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| These actions satisfy NEI 06-09-A, Section 2.3.5, Item 9.3.
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| E7 - 3
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| GNRO-2023/00014 Page 4 of 4 3.0 References
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| : 1. Letter from the NRC to NEI, Final Safety Evaluation for Nuclear Energy Institute (NEI)
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| Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (TAC No. MD4995), dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| : 2. Nuclear Energy Institute (NEI) Topical Report NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, dated October 2012 (ADAMS Accession No. ML12286A322)
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| : 3. Entergy Nuclear Management Manual EN-DC-151, PRA Maintenance and Update, Revision 9 E7 - 4
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| | |
| Enclosure 8 Grand Gulf Nuclear Station License Amendment Request 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 Model (4 Pages to Follow)
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| | |
| GNRO-2023/00014 Page 1 of 4 Attributes of the Real-Time Model 1.0 Introduction Section 4.0, Item 9 of the NRC Final Safety Evaluation (Reference 1) for NEI 06-09-A (Reference 2), requires that the license amendment request (LAR) provide a description of the PRA models and tools used to support the Risk-Managed Technical Specifications (RMTS).
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| This includes identification of how the baseline PRA models will be 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 models and CRMP tools, and training and qualification programs applicable to personnel responsible for development and use of the CRMP tools. This item also requires confirmation that the CRMP tools can be readily applied for each Technical Specification (TS) Limiting Condition for Operation (LCO) within the scope of the plant-specific submittal.
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| This enclosure describes the necessary changes to the peer-reviewed baseline PRA models for use in the CRMP software, which supports the Risk-Informed Completion Time (RICT) program.
| |
| The process that will be employed to adapt the baseline models is demonstrated:
| |
| : 1. to preserve the core damage frequency (CDF) and large early release frequency (LERF) quantitative results;
| |
| : 2. to maintain the quality of the peer-reviewed PRA models; and
| |
| : 3. to correctly accommodate changes in risk due to configuration-specific considerations.
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| Controls for ensuring PRA technical adequacy and training programs applicable to the RICT program are also discussed in this enclosure.
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| 2.0 Translation of Baseline Model for Use in Configuration Risk The baseline PRA models for internal events (including internal flooding) and internal fire are peer-reviewed models. These models are updated when necessary to incorporate changes to reflect the as-built, as-operated plant as discussed in Enclosure 7. The application specific model used for the RICT program will be a zero-maintenance model capable of calculating the internal events, internal flooding, and fire risk for each plant configuration. These models will be merged into a single one-top model for use in the CRMP. The CRMP model, which is used to implement the RICT program, will be verified to provide results equivalent to the baseline models in accordance with approved procedures.
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| The CRMP software will be used to facilitate all configuration-specific risk calculations and support RICT program implementation. The baseline PRA models are modified as follows for use in configuration risk calculations:
| |
| The unit availability factor is set to 1.0 (unit available).
| |
| * Maintenance unavailability is set to zero/false unless unavailable due to the actual
| |
| * configuration.
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| E8 - 1
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| GNRO-2023/00014 Page 2 of 4
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| * Mutually exclusive combinations, including normally disallowed maintenance combinations, are adjusted to allow accurate analysis of the configuration.
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| * For systems where some trains are in service and some in standby, the CRMP model will address the configurations of the plant in a manner to include defining in-service trains and alignments as needed.
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| * The impact of outside air temperatures on system requirements is addressed in the CRMP model. There are no changes in the PRA success criteria based on the time in the core operating cycle.
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| * The CRMP model accounts for severe weather conditions.
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| The CRMP software is designed to quantify the risk from internal events (including internal flooding) and fire and then use those values to calculate the risk management action time (RMAT) and RICT for each configuration. The RMAT and RICT will also account for the risk from seismic activity by applying a conservative CDF and LERF penalty to each calculation.
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| The treatment of common cause failure (CCF) will be in accordance with the approach described in NEI 06-09. For planned RICTs (e.g., to perform preventive maintenance tasks), no changes in CCF factors would be made in the CRMP model since no failures have occurred.
| |
| For emergent failures, Operations personnel would perform an extent of condition evaluation (using existing plant processes) to determine if any CCF potential exists. If it is determined that a CCF mode impacts redundant equipment and results in a TS loss of function, then the RICT program would not apply. If a potential CCF is determined to not result in a loss of function, then a quantitative or qualitative evaluation of the impact on the RICT would be performed. If a quantitative evaluation of increased CCF probability is performed, the adjustment to CCF probability will be made in accordance with Regulatory Guide 1.177, Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, Revision 2 (Reference 3). If a qualitative evaluation is performed, Risk Management Actions (RMAs) to manage a possible CCF would be considered for implementation. See Enclosure 12 for additional information regarding the development of RMAs with consideration of CCF modes.
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| 3.0 Quality Requirement and Consistency of PRA Model and Configuration Risk Management Tools The approach for establishing and maintaining the technical adequacy of the PRA models, including the CRMP model, includes both a PRA maintenance and update process (described in ), and the use of self-assessments and independent peer reviews (described in ).
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| The information provided in Enclosure 2 demonstrates that the Grand Gulf Nuclear Station internal events (including internal flooding), and fire PRA models conform to the associated industry standards endorsed by Regulatory Guide (RG) 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2 (Reference 4). This information provides a robust basis for concluding that the PRA models are technically adequate for use in risk-informed licensing actions.
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| For maintenance of an existing CRMP model, changes made to the baseline PRA model in translation to the CRMP model will be controlled and documented in accordance with Entergys E8 - 2
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| | |
| GNRO-2023/00014 Page 3 of 4 PRA configuration control procedure (Reference 5). The process will ensure models are accurate, as described in Enclosure 7. Because the CRMP model is developed from the complete baseline PRA models (i.e., it is not a simplified model), the results of this model would be expected to be essentially identical to those of the constituent baseline PRA models for internal events, internal flooding, and fire hazards (after accounting for the items in Section 2.0 such as setting maintenance events to zero, unit availability factor to 1.0, etc.). Acceptance testing will be performed after every CRMP model update to ensure that the software functions as intended and that quantification results are reasonable.
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| These actions satisfy NEI 06-09-A, Section 2.3.5, Item 9.
| |
| 4.0 Training and Qualification The PRA staff is responsible for development and maintenance of the CRMP model. Operations and Work Control staff will use the configuration risk tool to implement the RICT program. The PRA and Operations staff are trained in accordance with a program using National Academy for Nuclear Training ACAD documents, which is also accredited by the Institute of Nuclear Power Operations (INPO).
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| 5.0 Application of the Configuration Risk Tool to the RICT Program Scope The Electric Power Research Institute (EPRI) PHOENIX software, or equivalent, will be used to facilitate configuration-specific risk calculations and support the RICT program implementation.
| |
| This software is specifically designed to support the implementation of RMTS. The PHOENIX software will permit the user to evaluate all plant configurations using appropriate mapping of plant equipment to the PRA basic events. The equipment in the scope of the RICT program will be able to be evaluated in the appropriate PRA models. See Enclosure 1 for additional information regarding PRA functions and corresponding LCO Required Actions.
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| 6.0 References
| |
| : 1. Letter from the NRC to 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 (TAC No. MD4995), dated May 17, 2007 (ADAMS Accession No. ML071200238)
| |
| : 2. NEI Topical Report NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, dated October 2012 (ADAMS Accession No. ML12286A322)
| |
| : 3. NRC Regulatory Guide 1.177, Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, Revision 2, dated January 2021 (ADAMS Accession No. ML20164A034)
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| E8 - 3
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| GNRO-2023/00014 Page 4 of 4
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| : 4. NRC 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)
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| : 5. Entergy Nuclear Management Manual EN-DC-151, PRA Maintenance and Update, Revision 9 E8 - 4
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| Enclosure 9 GRAND GULF NUCLEAR STATION License Amendment Request 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 (4 Pages to Follow)
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| GNRO-2023/00014 Page 1 of 4 Key Assumptions and Sources of Uncertainty 1.0 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.
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| The NRC final Safety Evaluation for NEI Topical Report NEI 06-09-A (Reference 1), Section 4.0, Item 10 requires a discussion of how the key assumptions and sources of uncertainty were identified, and how their impact on the RMTS was assessed and dispositioned, such as through Risk Management Actions (RMA). The baseline internal events PRA (including internal flood) and fire PRA models document assumptions and sources of uncertainty and these were reviewed during the model peer reviews. Therefore, the approach taken is to review these documents to identify the items that 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.
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| The epistemic uncertainty analysis approach described below applies to the internal events (including internal flooding) PRA and any epistemic uncertainty impacts that are unique to the fire PRA are also addressed. In addition, NEI 06-09-A requires that the uncertainty be addressed in RICT Program Real Time Risk tools by consideration of the translation from the PRA model. The Real Time Risk model, also referred to as the Configuration Risk Management Program (CRMP) model, discussed in Enclosure 8 of this license amendment request (LAR),
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| will include internal events, internal 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 CRMP tool during RICT Program calculations.
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| The process used to evaluate sources of uncertainty for the RICT application follows the guidance illustrated in Figure 4-1 of EPRI TR-1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments (Reference 2). The sources of uncertainty evaluation for the baseline internal events PRA considers both plant-specific sources of uncertainty and the generic uncertainties identified in EPRI TR-1016737. The sources of uncertainty notebooks for the PRA models of record were reviewed to collect a listing of all sources of uncertainty that were identified as having potential impacts on the base PRA model or risk-informed applications. If the sources of uncertainty notebook already provided a justification that the model uncertainty need not be evaluated further as a potential source of uncertainty for the base model or for applications (e.g., negligible contribution, best-estimate modeling, etc.), then those model uncertainties were not considered further. This information represents the input from the base model assessment as shown in Figure 4-1. Application-specific uncertainties, as shown in Figure 4-1 of EPRI TR-1016737, are addressed in Section 3.0 of this enclosure.
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| 2.0 Assessment of Internal Events and Fire PRA Epistemic Uncertainty Impacts The calculation of a RICT is based on Incremental Core Damage Probability (ICDP) and Incremental Large Early Release Probability (ILERP). These are delta-risk measures that E9 - 1
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| GNRO-2023/00014 Page 2 of 4 evaluate the change in risk over the baseline zero maintenance risk for the plant. In reviewing each of the candidate sources of uncertainty for the internal events/internal flooding and fire PRAs, the following considerations were applied to determine if a RICT impact could exist. If a candidate source of uncertainty could be shown to satisfy these considerations, then this was considered to be adequate.
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| * Criterion #1: Candidate uncertainties that are qualitatively shown to have a very small impact on total risk and would be expected to have a negligible impact on delta-CDF and delta-LERF (particularly uncertainties that pertain to parts of the model that would not impact components that are in the RICT program, such as changes to non-support system initiating event frequencies, human error probabilities not related to RICT-eligible equipment, etc.).
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| * Criterion #2: Candidate uncertainties that are represented through conservative PRA modeling that would be expected to have a negligible or conservative impact on delta-risk RICT calculations.
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| * Criterion #3: Candidate uncertainties that were identified, but for which current industry-accepted approaches and data were used, are not considered as key sources of uncertainty. This is consistent with the ASME/ANS PRA Standard definition of a source of modeling uncertainty which states: a source is related to an issue in which there is no consensus approach or model and where the choice of approach or model is known to have an effect on the PRA model. A number of these candidates were derived from the EPRI list of generic PRA uncertainties.
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| * Criterion #4: Candidate uncertainties that were examined via sensitivity studies to confirm that the impact on baseline CDF and LERF are negligibly small are not considered as key sources of uncertainty for the RICT program.
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| There were nine assumptions and uncertainties identified from the internal events/internal flood PRA model that could potentially impact the RICT calculations. After review of each was completed, there was one source of uncertainty identified that could significantly impact the RICT calculations. This uncertainty topic is related to conservative modeling assumptions with the loss of decay heat removal contribution to LERF. A RICT specific sensitivity was performed for this source of uncertainty. The sensitivity study showed a significant increase in RICT as a function of LERF. However, the relevant LCOs are bounded by CDF and therefore, this source of uncertainty has no impact on the RICT calculations. For the fire PRA model, there were six assumptions and uncertainties identified that could potentially impact the RICT calculations.
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| After review of each uncertainty was completed, it was determined that none of the assumptions or uncertainties could impact the RICT calculations in a non-conservative manner. One of the assumptions reviewed for the fire PRA model involved Diverse and Flexible Coping Strategies (FLEX). The fire PRA takes very minimal credit for FLEX strategies, which includes RCIC injection from the upper containment pool and powering the H2 igniters. This was concluded not to be a key source of uncertainty for the RICT application.
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| E9 - 2
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| GNRO-2023/00014 Page 3 of 4 3.0 Assessment of Translation (Real Time Risk Model) Uncertainty Impacts Incorporation of the baseline PRA models into the CRMP model used for RICT Program calculations may introduce new sources of model uncertainty. Table E9-1 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 Enclosure 8 for additional discussion on the CRMP model.
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| Table E9-1: Assessment of Translation Uncertainty Impacts CRMP Model Change Part of Model Impact on Model Disposition and Assumption Affected PRA model logic structure Fault tree logic The model, if Since the may be optimized to model structure, restructured, will be restructured model increase solution speed. affecting internal logically equivalent will produce events PRAs and produce results comparable comparable to the numerical results, base PRA logic this is not a source of model. uncertainty for the RICT program.
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| Incorporation of seismic Calculation of The addition of Since this is a risk bias to support RICT RICT and risk conservative conservative Program risk calculations. management impacts for seismic approach for Conservative values for action time events has no addressing seismic the seismic delta CDF and (RMAT) within impact on base PRA risk in the RICT LERF are applicable. the CRM model or CRM model. Program, it is not a Impact is reflected in source of translation calculation of all uncertainty, and RICTs and RMATs. RICT Program calculations are not impacted. Therefore, no mandatory Risk Management Actions (RMAs) are required.
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| E9 - 3
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| GNRO-2023/00014 Page 4 of 4 CRMP Model Change Part of Model Impact on Model Disposition and Assumption Affected Set plant availability Risk metric Initiating event This change is (Reactor Critical Years calculated in per frequencies are consistent with CRM Factor) basic event to 1.0. reactor critical calculated in per Tool practices; years versus per reactor critical years. therefore, this calendar years The availability change does not factor is used in the represent a source of base PRA to convert translation the risk metric to uncertainty and RICT calendar years for program calculations average risk. The are not impacted.
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| CRM model Therefore, no evaluates specific mandatory RMAs are configurations required.
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| during at-power conditions with the reactor critical, so the conversion is not required, and the factor is 1.0.
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| 4.0 References
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| : 1. Letter from the NRC to NEI, "Final Safety Evaluation for Nuclear Energy Institute (NEI)
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| Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines' (TAC No. MD4995)," dated May 17, 2007 (ADAMS Accession No. ML071200238).
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| : 2. EPRI Technical Report TR-1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments," dated December 2008.
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| E9 - 4
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| Enclosure 10 Grand Gulf Nuclear Station License Amendment Request Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b Program Implementation (3 Pages to Follow)
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| GNRO-2023/00014 0 Page 1 of 3 Program Implementation 1.0 Introduction Section 4.0, Item 11 of the NRC Final Safety Evaluation (Reference 1) for NEI 06-09-A (Reference 2) requires that the license amendment request (LAR) provide a description of the implementing programs and procedures regarding plant staff responsibilities for the Risk Managed Technical Specifications (RMTS) implementation, and specifically discuss the decision process for risk management action (RMA) implementation during a Risk-Informed Completion Time (RICT).
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| 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 (CT). Enclosure 12 provides additional details regarding the process for developing and implementing RMAs.
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| 2.0 RICT Program and Procedures Entergy will develop a program description and implementing procedures for the RICT program.
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| The program description will establish the management responsibilities and general requirements for risk management, training, implementation, and monitoring of the RICT program. More detailed procedures will provide specific responsibilities, limitations, and instructions for implementing the RICT program. The program description and implementing procedures will incorporate the programmatic requirements for RMTS in accordance with NEI 06-09-A. The program will be integrated with the online work control process. The work control process currently identifies the need to enter a Limiting Condition for Operation (LCO) action statement as part of the planning process and will additionally identify whether the provisions of the RICT program are requirements for the planned work. The risk thresholds associated with 10 CFR 50.65(a)(4) 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.
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| The Operations Department (licensed operators) is responsible for compliance with the Technical Specification (TS) and will be responsible for the implementation of the 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.
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| The procedures for the RICT program will address the following attributes consistent with NEI 06-09-A:
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| * Plant management positions with authority to approve entry into the RICT program.
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| * Important definitions related to the RICT program.
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| * Departmental and position responsibilities for activities in the RICT program.
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| E10 - 1
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| GNRO-2023/00014 0 Page 2 of 3
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| * Plant conditions for which the RICT program is applicable.
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| * Limitations on implementing RICTs under voluntary and emergent conditions.
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| * Implementation of the RICT program 30-day back-stop CT limit.
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| * Use of the Configuration Risk Management Program (CRMP) tool.
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| * Implementation of the RICT and risk management action time (RMAT) within 12 hours or within the most limiting front-stop CT after a plant configuration change.
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| * Requirement 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.
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| * Guidance on the use of RMAs including the conditions under which they may be credited in RICT calculations.
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| * Guidance on crediting PRA functionality.
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| * Conditions for exiting a RICT.
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| * Requirements for training on the RICT program.
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| * Documentation requirements related to individual RICT evaluations, implementation of extended CTs, and accumulated annual risk.
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| 3.0 RICT Program Training Training will be carried out in accordance with Entergy training procedures and processes that utilize the Systematic Approach to Training. These procedures were written based on the Institute of Nuclear Power Operations (INPO) Accreditation requirements, as developed and maintained by the National Academy for Nuclear Training.
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| Participation Departments that will receive training appropriate to their level of program responsibilities include:
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| * Operations
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| * Operations Training
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| * Work Management
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| * Outage Management
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| * Planning and Scheduling Personnel
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| * Work Week Managers
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| * Regulatory Assurance
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| * Maintenance
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| * Engineering
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| * PRA Staff
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| * Other Selected Management Scope of Training For those individuals who will be directly involved in the implementation of the RICT program, the training will include the following topics:
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| E10 - 2
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| GNRO-2023/00014 0 Page 3 of 3
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| * Specific training on the revised TS
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| * Record keeping requirements
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| * Case studies
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| * Hands-on experience with the CRMP tool for calculating RMAT and RICT
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| * Identifying appropriate RMAs
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| * Determining PRA functionality
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| * Common cause failure considerations
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| * Other detailed programmatic requirements of the RICT program For management positions with authority to approve entry into the RICT program, as well as supervisors, managers, and other personnel who will closely support RICT implementation or need a general awareness, the training will provide a broad understanding of the purpose, concepts, and limitations of the program.
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| 4.0 References
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| : 1. Letter from the NRC to NEI, Final Safety Evaluation for Nuclear Energy Institute (NEI)
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| Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (TAC No. MD4995), dated May 17, 2007 (ADAMS Accession No. ML071200238)
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| : 2. Nuclear Energy Institute (NEI) Topical Report NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, dated October 2012 (ADAMS Accession No. ML12286A322)
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| E10 - 3
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| Enclosure 11 Grand Gulf Nuclear Station License Amendment Request Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b Monitoring Program (3 Pages to Follow)
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| GNRO-2023/00014 1 Page 1 of 3 Monitoring Program 1.0 Introduction Section 4.0, Item 12 of the NRC Final Safety Evaluation (Reference 1) for NEI 06-09-A (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, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 1 (Reference 3), and NEI 06-09-A. Note that Revision 2 of RG 1.174 (Reference 4) was issued by the NRC in May 2011 and made editorial changes to the applicable section of RG 1.174 referenced in Section 4.0, Item 12 of the NRC Safety Evaluation. Also, Revision 3 of RG 1.174 (Reference 5) was issued by the NRC in January 2018 and the relevant guidance regarding the implementation and monitoring program remain substantially the same.
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| 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). Calculation of the cumulative risk for the RICT program is discussed in Step 14 of Section 2.3.1 and Step 7.1 of Section 2.3.2 of NEI 06-09-A (Reference 2). General guidance for a performance monitoring program for risk-informed applications is discussed in Element 3 of RG 1.174, Revision 3 (Reference 5).
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| 2.0 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-A (Reference 2). 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). The total risk impact 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.,
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| beyond the front-stop CT). The change in risk will be converted to average annual values and documented every fuel cycle.
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| The total average annual change in risk for extended CTs will be compared to the guidance of RG 1.174, Revision 3 (Reference 5), Figures 4 and 5, 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.
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| The evaluation of the cumulative risk will also identify areas for consideration, such as:
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| * RICT applications that dominated and incurred a large portion of the risk increase E11 - 1
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| GNRO-2023/00014 1 Page 2 of 3
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| * Risk contributions from planned versus emergent RICT applications
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| * Risk Management Actions (RMAs) implemented but not credited in the risk calculations
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| * Risk impact from applying RICT to avoid multiple shorter duration outages Based on a review of the considerations above, corrective actions will be developed and implemented as appropriate. These actions may include:
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| * Administrative restrictions of the use of RICTs for specific high-risk configurations
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| * Additional RMAs for specific configurations
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| * Rescheduling planned maintenance activities
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| * Deferring planned maintenance to shutdown conditions
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| * Use of temporary equipment to replace out-of-service systems, structures, or components (SSCs)
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| * Plant modifications to reduce the risk impact of future planned maintenance configurations In addition to impacting cumulative risk, the implementation of the RICT program may potentially impact the unavailability of SSCs. The Maintenance Rule (MR) monitoring programs under 10 CFR 50.65 provide for evaluation and disposition of unavailability impacts that may be incurred from implementation of the RICT program. The MR program can be utilized for monitoring the performance of those SSCs that are in the scope of the RICT program and the scope of the MR.
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| The Grand Gulf Nuclear Station MR program structure is based on NUMARC 93-01 (Reference 6).
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| The monitoring program of the MR, along with the specific assessment of cumulative risk impact described above, serve as the RICT program implementation and monitoring program as described in Element 3 of RG 1.174 (Reference 5) and NEI 06-09-A (Reference 2).
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| 3.0 References
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| : 1. Letter from the NRC to NEI, Final Safety Evaluation for Nuclear Energy Institute (NEI)
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| Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (TAC No. MD4995), dated May 17, 2007 (ADAMS Accession No. ML071200238)
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| : 2. Nuclear Energy Institute (NEI) Topical Report NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, dated October 2012 (ADAMS Accession No. ML12286A322)
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| : 3. NRC 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)
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| : 4. NRC Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 2, dated May 2011 (ADAMS Accession No. ML100910006)
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| E11 - 2
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| GNRO-2023/00014 1 Page 3 of 3
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| : 5. NRC 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)
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| : 6. NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 4F, dated April 2018 (ADAMS Accession No. ML18120A069)
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| E11 - 3
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| Enclosure 12 Grand Gulf Nuclear Station License Amendment Request 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 (11 Pages to Follow)
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| GNRO-2023/00014 2 Page 1 of 11 Risk Management Action Examples 1.0 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 how the process would be implemented. Entergy procedures for planning and scheduling maintenance activities will govern RMAs. 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 NEI 06-09-A (Reference 1).
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| 2.0 Responsibilities For planned entries into the RICT program, Work Management is responsible for developing the RMAs with assistance from Operations and the PRA Staff. For emergent entry into extended CTs, Operations is responsible for developing the RMAs, but may seek assistance from the PRA Staff or Work Management. Operations is responsible for approval and implementation of all RMAs (for planned and emergent conditions).
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| 3.0 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 an RMAT already in place, the procedure will require a reevaluation of the existing RMAs for the new plant configuration to determine if additional RMAs are appropriate.
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| These requirements of the RICT program are consistent with the guidance of NEI 06-09-A.
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| 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) will be developed and implemented to address the potential for common cause failure modes. These RMAs will focus on reducing the likelihood of initiating events that rely on the affected function as well as protecting the in-service equipment that performs the redundant and diverse functions.
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| RMAs will be implemented no later than the time at which an incremental core damage probability (ICDP) of 1E-06 is reached, or no later than the time when an incremental large early release probability (ILERP) of 1E-07 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-03 per year or 1E-04 per year, respectively, RMAs are also required to be implemented. These requirements are consistent with the guidelines of NEI 06--
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| 09-A.
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| E12 - 1
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| GNRO-2023/00014 2 Page 2 of 11 By determining which initiators, fire compartments, or components are most important from a CDF or LERF perspective for a specific plant configuration, RMAs may be created to protect these components or increase awareness as it relates to their importance. Similarly, knowledge of the initiating event or sequence contribution related to the configuration-specific CDF or LERF allows development of RMAs that enhance the capability to mitigate such events. The guidance in NUREG-1855 (Reference 2) and EPRI Technical Update 1026511 (Reference 3) will be used in examining 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, important support systems, important human actions).
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| If the planned activity or emergent condition includes an SSC that is identified to impact fire PRA, as identified in the Configuration Risk Management Program (CRMP), fire PRA specific RMAs associated with that SSC will be implemented per procedure. Approved equipment-specific RMAs for risk significant SSCs within the scope of the RICT program will be contained in the procedure. Common cause RMAs will also be considered for emergent conditions where the extent of condition cannot rule out the potential for common cause failures.
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| 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-A. Nonetheless, if RMAs will be credited in RICT calculations, procedure instructions will be consistent with the guidance in NEI 06-09-A.
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| NEI 06-09-A classifies RMAs into the three categories described below:
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| : 1) Actions to increase risk awareness and control.
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| * Conduct shift briefings
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| * Conduct pre-job briefings
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| * Require the presence of the system engineer or other expertise related to the activity
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| * Receive management approval of the proposed activity
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| * Prioritize the restoration of out-of-service components
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| * Identify and protect important in-service components
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| : 2) Actions to reduce the duration of maintenance activities.
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| * Pre-stage parts and materials
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| * Perform walk-downs of the system tag-outs and key equipment prior to beginning the work
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| * Develop critical activity procedures for risk-significant configurations, including identification of the associated risk and contingency plans for approaching/exceeding the RICT
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| * Conduct training on mock-ups to familiarize maintenance personnel prior to the work
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| * Perform the activity around the clock rather than day-shift only
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| * Establish contingency plans to restore key out-of-service equipment rapidly if needed
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| : 3) Actions to minimize the magnitude of the risk increase.
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| * Suspend or minimize activities on or in the vicinity of redundant systems
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| * Suspend or minimize activities on other systems that adversely affect the CDF or LERF E12 - 2
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| GNRO-2023/00014 2 Page 3 of 11
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| * 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 designed to mitigate
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| * Use temporary equipment to provide backup power, ventilation, etc.
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| * Reschedule other risk-significant activities
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| * Expedite equipment return to service to reduce risk levels
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| * Protect equipment that provides alternate diverse success paths for performing the safety function of the out-of-service SSC Determining RMAs involves the use of both qualitative and quantitative considerations for the specific plant configuration and the practical means available to manage risk. A graded approach is used to identify the scope of RMAs that are appropriate for the plant configuration.
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| Procedural guidance for the development of RMAs in support of the RICT program builds off other processes, such as actions taken under the 10 CFR 50.65(a)(4) program and the protected equipment program. Additionally, common cause RMAs may be developed to address the potential impact of common cause failure modes and may be performed in conjunction with the Grand Gulf Nuclear Station Safety Function Determination Program, per Technical Specification (TS) 5.5.10.
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| Entergy procedures will provide general guidance for developing RMAs, such as:
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| * Consideration of rescheduling maintenance to reduce risk
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| * Discussion of RICT in pre-job briefs
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| * Consideration of proactive return-to-service of other equipment
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| * Efficient execution of the maintenance In addition to RMAs developed qualitatively, RMAs are developed based on the CRMP to identify configuration-specific RMA candidates to manage the risk associated with internal events, internal flooding, and fire events. These actions include:
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| * Identification of important equipment or divisions for protection
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| * Identification of important operator actions for crew briefings or just-in-time training
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| * Identification of key fire initiators and fire zones for RMAs
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| * Identification of dominant initiating events and actions to minimize their occurrence
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| * Consideration of insights from dominant PRA model cutsets Common cause RMAs may also be developed to ensure availability of redundant SSCs, to ensure availability of diverse systems, to reduce the likelihood of initiating events that the out-of-service components are designed to mitigate, 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.
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| Examples of common cause RMAs include:
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| * Performance of non-intrusive inspections on alternate divisions
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| * More frequent monitoring for running or standby components
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| * Expansion of component monitoring E12 - 3
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| GNRO-2023/00014 2 Page 4 of 11
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| * Deferring maintenance and testing activities that could generate or increase the frequency of an initiating event which would require operation of potentially affected SSCs
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| * Readiness of operators and maintenance to respond to additional failures
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| * Shift briefs or standing orders that focus on initiating event response or loss of potentially affected SSCs
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| * Engineering evaluation of a failed component to determine the most likely cause and to evaluate the susceptibility of other plant equipment Entergy procedures will require that for emergent conditions where the extent of condition is not performed prior to exceeding the RMAT or the extent of condition cannot rule out the potential for common cause failure, common cause RMAs are expected to be implemented. These RMAs can include the pre-identified, general RMAs as discussed above, as well as alternative common cause RMAs for the specific configuration.
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| 4.0 Examples Multiple example 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 the inoperability of a Diesel Generator (DG), an offsite electrical power circuit, an offsite circuit with a DG, a safety-related 125V DC battery, an electrical distribution subsystem, and a low-pressure Emergency Core Cooling System (ECCS) injection/spray subsystem.
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| 4.1 DG Inoperable For TS 3.8.1.B, one required DG inoperable, RMAs may include:
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| : 1) Actions to increase risk awareness and control.
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| * Brief the on-shift operations crew concerning the unit activities, including any compensatory measures established. Specific focus areas may be to review emergency or abnormal operating procedures associated with loss of offsite power (LOOP) and station blackout (SBO) events. This may include a review of alternate electrical alignments, considering the inoperable DG, that may be needed during a LOOP or SBO event.
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| * Perform walkdowns of the remaining operable DGs to validate their standby/readiness condition.
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| * Communicate the configuration to the transmission system operator so that any planned activities with the potential to cause a grid disturbance can be closely coordinated or deferred.
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| * Minimize the accumulation of transient combustibles in accordance with the station fire protection program; this would include a heightened sense of awareness on the fire zones that have become more important due to the inoperable DG.
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| * Challenge any plans to conduct hot work inside the power block or in the switchyard and particularly in the fire zones that have become more important due to the configuration. There should be a bias for not performing hot work during the extended CT (unless its directly related to the inoperable DG).
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| E12 - 4
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| GNRO-2023/00014 2 Page 5 of 11
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| : 2) Actions to reduce the duration of maintenance activities.
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| * For preplanned RICT entry, use an equipment outage schedule that identifies and plans all needed resources and provides logic ties between critical activities.
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| * Confirm parts availability prior to entry into a preplanned RICT.
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| * Designate additional resources to improve efficiency (e.g., designate a resource to be a parts or tools expediter to maximize wrench time).
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| * Walkdown systems prior to beginning the work and stage equipment (hoses, fittings, tools, etc.) needed to conduct the tagging and the maintenance.
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| * Anticipate and be prepared to tag-in the DG as the maintenance concludes (be prepared to refill lube oil systems, fill and vent coolant water, rack breakers, etc.).
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| * Ensure the correct post-maintenance DG testing is planned and is ready to be implemented following the maintenance (operators are briefed, procedures are ready, etc.).
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| * Procure the DG vendors services and have them on site to support the equipment outage, if needed.
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| : 3) Actions to minimize the magnitude of the risk increase.
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| * Verify the availability of required offsite circuits every eight hours.
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| * Ensure safety functions are maintained by continually assessing the status of required features that are redundant to those systems supported by the inoperable DG.
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| * For emergent RICT entry, determine if the remaining operable DGs are impacted by a common cause failure mode.
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| * Proactively implement RMAs during times of high grid stress conditions, such as during high demand conditions.
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| * Evaluate weather conditions for threats to the reliability of offsite power supplies.
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| * Defer elective maintenance in the switchyard and on the stations electrical distribution systems.
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| * Defer maintenance or testing that may impact the reliability of operable DGs and associated support equipment.
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| * Implement 10 CFR 50.65(a)(4) fire-specific RMAs associated with the affected DG.
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| * Implement the protected equipment program for the inoperable DG (protect operable DGs and other important equipment identified in the CRMP).
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| * Maintain detection, suppression, and fire zone barriers intact and minimize transient combustibles for those fire areas identified as being significant for the configuration.
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| * Curtail non-essential electrical switching operations to minimize the potential for deenergizing electrical buses.
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| 4.2 Offsite Circuit Inoperable For TS 3.8.1.A, one required offsite circuit inoperable, RMAs may include:
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| : 1) Actions to increase risk awareness and control.
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| * Brief the on-shift operations crew concerning the unit activities, including any compensatory measures established. Specific focus areas would be to review E12 - 5
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| | |
| GNRO-2023/00014 2 Page 6 of 11 emergency or abnormal operating procedures associated with LOOP and SBO events. This may include a review of alternate switchyard alignments considering the inoperable offsite source.
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| * Limit access (e.g., vehicle traffic) and other activities in the switchyard; perform periodic walkdowns of the switchyard to validate no unauthorized activities are in progress.
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| * Routinely communicate with the transmission system operator so that any planned activities with the potential to cause a grid disturbance can be closely coordinated or deferred.
| |
| * Minimize the accumulation of transient combustibles in accordance with the station fire protection program; this would include a heightened sense of awareness on the fire zones that have become more important due to the plant configuration (e.g.,
| |
| onsite power sources will become more important).
| |
| * Challenge any plans to conduct hot work inside the power block or in the switchyard and particularly in the fire zones that have become more important due to the configuration. There should be a bias for not performing hot work during the extended CT (unless its directly related to the inoperable offsite circuit).
| |
| : 2) Actions to reduce the duration of maintenance activities.
| |
| * For preplanned RICT entry, use an equipment outage schedule that identifies and plans all needed resources and provides logic ties between critical activities.
| |
| * Confirm parts availability prior to entry into a preplanned RICT.
| |
| * Implement rigorous communications and coordination between site operations and transmission grid operators to ensure the activities rendering the offsite circuit inoperable are progressing as expected and if not, additional resources are applied to reduce the duration of the activity.
| |
| : 3) Actions to minimize the magnitude of the risk increase.
| |
| * Verify the availability of the remaining offsite circuit every eight hours.
| |
| * Proactively consider additional RMAs during times of high grid stress conditions, such as during high demand conditions.
| |
| * Evaluate weather conditions for threats to the reliability of the remaining offsite power source.
| |
| * Defer elective maintenance in the switchyard and on the stations electrical distribution systems.
| |
| * Defer maintenance or testing that may impact the reliability of DGs and associated support equipment.
| |
| * Implement 10 CFR 50.65(a)(4) fire-specific RMAs associated with the offsite source.
| |
| * Implement the protected equipment program for the inoperable offsite source. This may include limiting access to the switchyard (e.g., locked gates), barricading and/or posting transformer areas, and/or posting rooms containing intermediate voltage electrical buses.
| |
| * Maintain detection, suppression, and fire zone barriers intact and minimize transient combustibles for those fire areas identified as being significant for the configuration.
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| E12 - 6
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| GNRO-2023/00014 2 Page 7 of 11
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| * Curtail non-essential electrical switching operations to prevent losing the remaining offsite source and to prevent challenging the availability of onsite electrical power.
| |
| 4.3 Offsite Circuit Inoperable with an Inoperable DG For TS 3.8.1.D, one required offsite circuit inoperable and one required DG inoperable, RMAs may include:
| |
| : 1) Actions to increase risk awareness and control.
| |
| * Brief the on-shift operations crew concerning the unit activities, including any compensatory measures established. Specific focus areas would be to review emergency or abnormal operating procedures associated with LOOP and SBO events. This may include a review of alternate electrical alignments.
| |
| * Limit access (e.g., vehicle traffic) and other activities in the switchyard; perform periodic walkdowns of the switchyard to validate no unauthorized activities are in progress.
| |
| * Communicate with the transmission system operator so that any planned activities with the potential to cause a grid disturbance can be closely coordinated or deferred.
| |
| In this condition there should be a bias towards deferral.
| |
| * Minimize the accumulation of transient combustibles in accordance with the station fire protection program; this would include a heightened sense of awareness on the fire zones that have become more important due to the plant configuration.
| |
| * Perform walkdowns of the remaining operable DGs to validate their standby/readiness condition.
| |
| * Challenge any plans to conduct hot work inside the power block or in the switchyard and particularly in the fire zones that have become more important due to the configuration. There should be a bias for not performing hot work during the extended CT (unless its directly related to the inoperable offsite circuit or the inoperable DG).
| |
| * Establish the outage control center to provide oversight and coordinate return-to-service activities.
| |
| : 2) Actions to reduce the duration of maintenance activities.
| |
| * Implement rigorous communications and coordination between site operations and transmission grid operators to ensure the activities rendering the offsite circuit inoperable are progressing as expected and if not, additional resources are applied to reduce the duration of the activity.
| |
| * Engage senior managers and executives to ensure the transmission operators are utilizing the necessary resources with an extreme sense of urgency as it relates to restoring the offsite circuit to an operable status (considering the concurrent DG inoperability).
| |
| * Anticipate and be prepared to tag-in the DG as the maintenance concludes (be prepared to refill lube oil systems, fill and vent coolant water, rack breakers, etc.).
| |
| * Ensure the correct post-maintenance DG testing is planned and is ready to be implemented following the maintenance (operators are briefed, procedures are ready, etc.).
| |
| E12 - 7
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| GNRO-2023/00014 2 Page 8 of 11
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| * Procure the DG vendors services and have them on site to support the equipment outage, if needed.
| |
| : 3) Actions to minimize the magnitude of the risk increase.
| |
| * Verify the availability of the remaining offsite circuit every eight hours.
| |
| * Ensure safety functions are maintained by continually assessing the status of required features that are redundant to those systems supported by the inoperable DG.
| |
| * Determine if the remaining operable DGs are impacted by a common cause failure mode.
| |
| * Implement additional RMAs during times of high grid stress conditions, such as during high demand conditions.
| |
| * Evaluate weather conditions for threats to the reliability of the remaining offsite power supplies.
| |
| * Allow only mission critical maintenance (no activities that arent related to restoring the DG or the offsite source) in the switchyard or on the stations electrical distribution systems.
| |
| * Defer maintenance or testing that may impact the reliability of DGs and associated support equipment.
| |
| * Implement 10 CFR 50.65(a)(4) fire-specific RMAs associated with the offsite source and the inoperable DG.
| |
| * Implement the protected equipment program for the inoperable offsite source and the inoperable DG. This may include limiting access to the switchyard (e.g., locked gates), barricading and/or posting transformer areas, and posting rooms containing intermediate voltage electrical buses, etc.
| |
| * Maintain detection, suppression, and fire zone barriers intact and minimize transient combustibles for those fire areas identified as being significant for the configuration.
| |
| * Curtail non-essential electrical switching operations to prevent losing the remaining offsite source and to prevent challenging the availability of onsite electrical power.
| |
| 4.4 Safety-Related 125V DC Battery Inoperable For TS 3.8.4.C, one Division I or II DC electrical power subsystem inoperable, RMAs may include:
| |
| : 1) Actions to increase risk awareness and control.
| |
| * Brief the on-shift operations crew concerning the unit activities, including any compensatory measures established. Specific focus areas would be to review operating procedures associated with loss of DC power and SBO events.
| |
| * Brief the on-shift operations crew concerning the impact that the out-of-service battery would have on the potential response to plant events (e.g., reduced control systems).
| |
| * Minimize the accumulation of transient combustibles in accordance with the station fire protection program for the impacted fire zones.
| |
| * Minimize activities that could trip the unit (e.g., limit maintenance or testing on reactor protection system instrumentation).
| |
| E12 - 8
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| GNRO-2023/00014 2 Page 9 of 11
| |
| * Challenge any plans to conduct hot work inside the power block or in the switchyard and particularly in the fire zones that have become more important due to the configuration. There should be a bias for not performing hot work during the extended CT (unless its directly related to the inoperable battery).
| |
| : 2) Actions to reduce the duration of maintenance activities.
| |
| * Confirm parts availability prior to entry into a preplanned RICT.
| |
| * Walkdown systems prior to beginning the work.
| |
| * Anticipate and be prepared to tag-in the system as the maintenance concludes.
| |
| * Ensure the correct post-maintenance testing and other restoration activities are planned and are ready to be implemented following the maintenance (e.g., to equalize battery voltage with the bus).
| |
| : 3) Actions to minimize the magnitude of the risk increase.
| |
| * Proactively implement RMAs during times of high grid stress conditions, such as during high demand conditions.
| |
| * Evaluate weather conditions for threats to the reliability of offsite power supplies.
| |
| * Defer elective maintenance in the switchyard and on the stations electrical distribution systems.
| |
| * Defer maintenance or testing that affects the reliability of the DGs and their associated support equipment.
| |
| * Protect DC electrical buses, remaining operable DC batteries, and other support equipment.
| |
| * Implement 10 CFR 50.65(a)(4) fire-specific RMAs for the associated bus and battery.
| |
| * Maintain detection, suppression, and fire zone barriers intact and minimize transient combustibles for those fire zones identified as being risk significant for the configuration.
| |
| 4.5 Electrical Power Distribution Subsystems Inoperable For TS 3.8.7.A, one or more Division I or II AC electrical power distribution subsystem(s) inoperable, RMAs may include:
| |
| : 1) Actions to increase risk awareness and control.
| |
| * Brief the on-shift operations crew concerning the unit activities, including any compensatory measures established. Specific focus areas may be to review operating procedures and the electrical load list for the inoperable AC distribution subsystem.
| |
| * Brief the on-shift operations crew concerning the impact the AC power subsystem has on the potential response to plant events.
| |
| * Minimize the accumulation of transient combustibles in accordance with the station fire protection program for the impacted fire zones.
| |
| * Challenge hot work in important fire zones. There should be a bias for not performing hot work during the extended CT (unless its directly related to the inoperable AC electrical power subsystem).
| |
| E12 - 9
| |
| | |
| GNRO-2023/00014 2 Page 10 of 11
| |
| : 2) Actions to reduce the duration of maintenance activities.
| |
| * For preplanned RICT entry, use an equipment outage schedule that identifies and plans all needed resources and provides logic ties between critical activities.
| |
| * Confirm parts availability prior to entry into a preplanned RICT.
| |
| * Walkdown systems prior to beginning the work.
| |
| : 3) Actions to minimize the magnitude of the risk increase.
| |
| * Defer elective maintenance in the switchyard and on the stations electrical distribution systems.
| |
| * Defer maintenance or testing that affects the reliability of the DGs or their associated support equipment.
| |
| * Protect redundant AC electrical power subsystems and their support equipment.
| |
| * Implement 10 CFR 50.65(a)(4) fire-specific RMAs for the associated AC bus.
| |
| * Maintain detection, suppression, and fire zone barriers intact and minimize transient combustibles for those fire zones identified as being risk significant for the configuration.
| |
| 4.6 Low Pressure ECCS Injection/Spray Subsystem Inoperable For TS 3.5.1.A, one low pressure ECCS injection/spray subsystem inoperable, RMAs may include:
| |
| : 1) Actions to increase risk awareness and control.
| |
| * Brief the on-shift operations crew concerning the unit activities, including any compensatory measures established. Specific focus areas would be to review procedures associated with reactor coolant system (RCS) leakage or loss of coolant accidents.
| |
| * Perform a walkdown and validation of the remaining ECCS systems to validate the standby/readiness condition.
| |
| * Minimize the accumulation of transient combustibles in accordance with the station fire protection program for the impacted fire zones.
| |
| * Assess the RCS operational leakage rate prior to entering the RICT; consider deferral of a planned RICT if RCS leakage is elevated and/or the source of leakage is unknown.
| |
| : 2) Actions to reduce the duration of maintenance activities.
| |
| * For preplanned RICT entry, use an equipment outage schedule that identifies and plans all needed resources and provides logic ties between critical activities.
| |
| * Confirm parts availability prior to entry into a preplanned RICT.
| |
| * Walkdown systems prior to beginning the work.
| |
| * Anticipate and be prepared to tag-in the system as the maintenance concludes (be prepared to fill and vent systems, rack breakers, etc.)
| |
| * Ensure the correct post-maintenance testing is planned and is ready to be implemented following the maintenance (operators are briefed, procedures are ready, etc.)
| |
| E12 - 10
| |
| | |
| GNRO-2023/00014 2 Page 11 of 11
| |
| : 3) Actions to minimize the magnitude of the risk increase.
| |
| * Defer maintenance or testing activities on the redundant ECCS systems and associated support equipment and protect those systems (e.g., post doors to limit access).
| |
| * Defer maintenance or testing that affects the reliability of those safety systems that provide defense-in-depth. If testing or maintenance activities must be performed, a review of the potential risk impact should be performed.
| |
| * Minimize activities that could trip the unit or increase the frequency of an initiating event.
| |
| * Verify system alignment of remaining ECCS systems.
| |
| * Implement equipment protection for redundant components and diverse systems.
| |
| 5.0 References
| |
| : 1. Nuclear Energy Institute (NEI) Topical Report NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, dated October 2012 (ADAMS Accession No. ML12286A322)
| |
| : 2. NRC NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, Revision 1, dated March 2017 (ADAMS Accession No. ML17062A466)
| |
| : 3. EPRI Technical Update 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
| |
| : 4. Entergy Nuclear Management Manual EN-WM-104, On Line Risk Assessment, Revision 23 E12 - 11}}
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