RS-20-052, Application to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Extended Completed Times-RITSTF Initiative 4b: Difference between revisions

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{{#Wiki_filter:4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-20-052                                                                                    10 CFR 50.90 April 30, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461
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==Subject:==
Application to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License No. NPF-62 for Clinton Power Station (CPS), Unit 1. The proposed amendment would modify Technical Specifications (TS) requirements to permit the use of Risk Informed Completion Times (RICTs) in accordance with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," (ADAMS Accession No. ML18183A493). A model safety evaluation was provided by the NRC to the TSTF on November 21, 2018 (ADAMS Accession No. ML18267A259).
* Attachment 1 provides a description and assessment of the proposed changes, the requested confirmation of applicability, and plant-specific verifications.
* Attachment 2 provides the existing TS pages marked up to show the proposed changes.
* Attachment 3 provides the existing TS Bases pages marked up to show the proposed changes and is provided for information only.
* Attachment 4 provides a cross-reference between the Technical Specifications included in TSTF-505, Rev. 2 and the CPS plant-specific TS.
* Attachment 5 provides information supporting the redundant means available to mitigate accidents for instrumentation governed by the TS proposed to be included as part of the RICT program in this submittal.
The proposed change has been reviewed by the Plant Operations Review Committee in accordance with the requirements of the EGC Quality Assurance Program.
 
April 30, 2020 U.S. Nuclear Regulatory Commission Page 2 EGC requests approval of the proposed change by April 30, 2021. Once approved, the amendment will be implemented within 180 days. This implementation period will provide adequate time for the affected station documents to be revised using the appropriate change control mechanisms.
In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (a)(1), the analysis about the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is being provided to the Commission.
In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"
paragraph (b), EGC is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Kenneth M. Nicely at (630) 657-2803.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 30th day of April 2020.
Respectfully, Patrick R. Simpson Sr. Manager Licensing Attachments:
: 1. Description and Assessment
: 2. Proposed Technical Specification Changes - Mark-Ups
: 3. Proposed Technical Specification Bases Changes - Mark-Ups (For Information Only)
: 4. Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications
: 5. Information Supporting Instrumentation Redundancy and Diversity
 
April 30, 2020 U.S. Nuclear Regulatory Commission Page 3
 
==Enclosures:==
: 1. List of Revised Required Actions to Corresponding PRA Functions
: 2. Information Supporting Consistency with Regulatory Guide 1.200, Revision 2
: 3. Information Supporting Technical Adequacy of PRA Models Without PRA Standards Endorsed by Regulatory Guide 1.200, Revision 2
: 4. Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
: 5. Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF)
: 6. Justification of Application of At-Power PRA Models to Shutdown Modes
: 7. PRA Model Update Process
: 8. Attributes of the Real Time Risk Model
: 9. Key Assumptions and Sources of Uncertainty
: 10. Program Implementation
: 11. Monitoring Program
: 12. Risk Management Action Examples cc:    NRC Regional Administrator, Region III NRC Senior Resident Inspector - Clinton Power Station Illinois Emergency Management Agency - Division of Nuclear Safety
 
ATTACHMENT 1 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Description and Assessment
 
ATTACHMENT 1 Description and Assessment
 
==1.0      DESCRIPTION==
 
In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC), requests an amendment to Facility Operating License No. NPF-62 for Clinton Power Station (CPS), Unit 1.
The proposed amendment would modify the Technical Specifications (TS) requirements related to Completion Times (CTs) for Required Actions (Required Action allowed outage times for CPS) to provide the option to calculate a longer, risk-informed CT. A new program, the Risk-Informed Completion Time (RICT) Program, is added to TS Section 5.0, "Administrative Controls."
The methodology for using the RICT Program is described in NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)
Guidelines," Revision 0, which was approved by the NRC on May 17, 2007. Adherence to NEI 06-09-A is required by the RICT Program.
The proposed amendment is consistent with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b." However, only those Required Actions described in Attachment 4 and Enclosure 1, as reflected in the proposed TS mark-ups provided in Attachment 2, are proposed to be changed, because some of the modified Required Actions in TSTF-505 are not applicable to CPS, and there are some plant-specific Required Actions not included in TSTF-505 that are included in this proposed amendment.
2.0      ASSESSMENT 2.1      Applicability of Published Safety Evaluation EGC has reviewed TSTF-505, Revision 2, and the model safety evaluation dated November 21, 2018 (ADAMS Accession No. ML18267A259). This review included the information provided to support TSTF-505 and the safety evaluation for NEI 06-09-A. As described in the subsequent paragraphs, EGC has concluded that the technical basis is applicable to CPS, Unit 1, and support incorporation of this amendment in the CPS TS.
2.2      Verifications and Regulatory Commitments In accordance with Section 4.0, Limitations and Conditions, of the safety evaluation for NEI 06-09-A, the following is provided:
: 1. Enclosure 1 identifies each of the TS Required Actions to which the RICT Program will apply, with a comparison of the TS functions to the functions modeled in the probabilistic risk assessment (PRA) of the structures, systems and components (SSCs) subject to those actions.
: 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 required by Regulatory Guide (RG) 1.200, Section 4.2.
Page 1
 
ATTACHMENT 1 Description and Assessment
: 3. Enclosure 3 is not applicable since each PRA model used for the RICT Program is addressed using a standard endorsed by the NRC.
: 4. Enclosure 4 provides appropriate justification for excluding sources of risk not addressed by the PRA models.
: 5. Enclosure 5 provides the plant-specific baseline core damage frequency (CDF) and large early release frequency (LERF) to confirm that the potential risk increases allowed under the RICT Program are acceptable.
: 6. Enclosure 6 is not applicable since the RICT Program is not being applied to shutdown modes.
: 7. Enclosure 7 provides a discussion of the licensee's programs and procedures that assure the PRA models that support the RICT Program are maintained consistent with the as-built, as-operated plant.
: 8. Enclosure 8 provides a description of how the baseline PRA model, which calculates average annual risk, is evaluated and modified for use in the Real Time Risk tool to assess real time configuration risk, and describes the scope of, and quality controls applied to, the Real Time Risk tool.
: 9. Enclosure 9 provides a discussion of how the key assumptions and sources of uncertainty in the PRA models were identified, and how their impact on the RICT Program was assessed and dispositioned.
: 10. Enclosure 10 provides a description of the implementing programs and procedures regarding the plant staff responsibilities for the RICT Program implementation, including risk management action (RMA) implementation.
: 11. Enclosure 11 provides a description of the implementation and monitoring program as described in NEI 06-09-A, Section 2.3.2, Step 7.
: 12. Enclosure 12 provides a description of the process to identify and provide RMAs.
2.3    Optional Changes and Variations EGC is proposing the following variations from the TS changes described in TSTF-505, Revision 2, or the applicable parts of the NRC's model safety evaluation dated November 21, 2018. These options were recognized as acceptable variations in TSTF-505 and the NRC's model safety evaluation.
In a few instances, the CPS TS use different numbering and titles than the Standard Technical Specifications (STS) on which TSTF-505 was based. These differences are administrative and do not affect the applicability of TSTF-505 to the CPS TS. Attachment 4 provides specific information. is a cross-reference that provides a comparison between the NUREG-1434, "Standard Technical Specifications, General Electric BWR/6 Plants," Required Actions included Page 2
 
ATTACHMENT 1 Description and Assessment in TSTF-505 and the CPS Required Actions included in this license amendment request. The attachment includes a summary description of the referenced Required Actions, which is provided for information purposes only and is not intended to be a verbatim description of the Required Actions. The cross-reference in Attachment 4 identifies the following:
: 1. CPS 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 NRC's model safety evaluation dated November 21, 2018.
: 2. CPS Required Actions that have different numbering than the NUREG-1434 Required Actions are an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
: 3. For NUREG-1434 Required Actions that are not contained in the CPS TS, the corresponding TSTF-505 mark-ups for the Required Actions are not applicable to CPS.
This is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
: 4. The model application provided in TSTF-505 includes an attachment for typed, camera-ready (revised) TS pages reflecting the proposed changes. EGC is not including such an attachment due to the number of TS pages included in this submittal that have the potential to be affected by other unrelated license amendment requests and the straightforward nature of the proposed changes. Providing only mark-ups of the proposed TS changes satisfies the requirements of 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," in that the mark-ups fully describe the changes desired. This is an administrative deviation from TSTF-505 with no impact on the NRC's model safety evaluation dated November 21, 2018.
Because of this deviation, the contents and numbering of the attachments for this amendment request differ from the attachments specified in the model application in TSTF-505.
: 5. 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 CPS TS. CPS TS Section 1.2, "Logical Connectors," specifies that Completion Times only use first level logic. Therefore, the proposed mark-ups 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 CPS 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 NRC's model safety evaluation dated November 21, 2018.
: 6. There are several plant-specific LCOs and associated Required Actions for which EGC is proposing to apply the RICT Program that are variations from TSTF-505 as identified in Attachment 4. Attachment 4 was created using the BWR/6 standard from Page 3
 
ATTACHMENT 1 Description and Assessment NUREG-1434, with exceptions annotated on Attachment 4 and summarized below.
Additional details are contained in Attachment 4 for TS Conditions and Required Actions.
* TS 3.3.1.1 - Reactor Protection System (RPS) Instrumentation. The CPS RPS logic is different than the standard BWR/6 design upon which NUREG-1434 is based. For CPS, trip capability for each function is maintained with any two of four RPS channels operable. Additional discussion is provided in Attachments 4 and 5.
* TS 3.3.4.1 - End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation.
The CPS EOC-RPT logic is different than the standard BWR/6 design upon which NUREG-1434 is based. For CPS, trip capability for each function is maintained with any two of four EOC-RPT channels operable. Additional discussion is provided in Attachments 4 and 5.
* TS 3.3.6.1 - Primary Containment Isolation Instrumentation. The CPS MSL isolation logic is different than the standard BWR/6 design upon which NUREG-1434 is based. For CPS, trip capability for each function is maintained with any two of four MSL isolation channels operable. Additional discussion is provided in Attachments 4 and 5.
* TS 3.3.6.4 - Suppression Pool Makeup (SPMU) System Instrumentation.
TSTF-505 does not apply a RICT to TS 3.3.6.4. Industry guidance developed by the Technical Specifications Task Force indicates that Required Actions 3.3.6.4.B.2 and 3.3.6.4.C.2 were excluded because the traveler will not modify Required Actions that represent a loss of function. However, CPS proposes to apply a RICT to these actions because the risk impact can be assessed using a surrogate model. The proposed TS change includes a Note to indicate that the Risk Informed Completion Time Program is not applicable when trip capability is not maintained.
* TS 3.6.1.6 - Low-Low Set (LLS) Valves. TSTF-505 does not apply a RICT to TS 3.6.1.6, Required Action A.1. Industry guidance developed by the Technical Specifications Task Force indicates that this action was excluded because the traveler will not modify Required Actions for systems that do not affect CDF or LERF or for which a RICT cannot be quantitatively determined. However, CPS proposes to apply a RICT to this action because the function is modeled in the PRA and can be directly included in the RTR tool for the RICT program. As stated in the CPS TS Bases for Required Action A.1, with one LLS valve inoperable, the remaining operable LLS valves are adequate to perform the designed function.
* TS 3.8.9 - Distribution Systems - Operating. Required Action 3.8.9.B.1 is a plant-specific Condition with a Required Action to restore the Division 1 and 2 uninterruptible AC bus distribution subsystems to operable status. CPS proposes to apply a RICT to the existing CPS TS 3.8.9, Required Action B.1.
This is acceptable because the TSTF states that there may also be plant-specific TS to which changes of the type presented in the TSTF may be applied.
Page 4
 
ATTACHMENT 1 Description and Assessment EGC has determined that the application of a RICT for these CPS plant-specific LCOs is consistent with TSTF-505 and with the NRC's model safety evaluation dated November 21, 2018. Application of a RICT for plant-specific LCOs will be controlled under the RICT Program. The RICT Program provides the necessary administrative controls to permit extension of Completion Times and thereby delay reactor shutdown or remedial actions, if risk is assessed and managed within specified limits and programmatic requirements. The specified safety function or performance levels of TS required structures, systems or components (SSCs) are unchanged, and the remedial actions, including the requirement to shut down the reactor, are also unchanged; only the Required Action allowed outage times are extended by the RICT Program.
Application of a RICT will be evaluated using the methodology and probabilistic risk guidelines contained in NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," Revision 0, which was approved by the NRC on May 17, 2007 (ADAMS Accession No. ML071200238). 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 risk management actions (RMAs) to maintain the risk below acceptable regulatory risk thresholds. In addition, the NEI 06-09-A, Revision 0 methodology satisfies the five key safety principles specified in Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," dated August 1998 (ADAMS Accession No. ML003740176), relative to the risk impact due to the application of a RICT.
Therefore, the proposed application of RICT in the CPS plant-specific Required Actions is consistent with TSTF-505, Revision 2, and with the NRC's model safety evaluation dated November 21, 2018.
EGC has reviewed these changes and determined that they do not affect the applicability of TSTF-505, Revision 2, to the CPS TS.
 
==3.0    REGULATORY ANALYSIS==
 
3.1    No Significant Hazards Consideration Determination Exelon Generation Company, LLC (EGC) has evaluated the proposed changes to the Technical Specifications (TS) using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.
Clinton Power Station (CPS), Unit 1, requests adoption of an approved change to the standard technical specifications (STS) and plant-specific TS, to modify the TS requirements related to Completion Times for Required Actions to provide the option to calculate a longer, risk-informed Completion Time. The allowance is described in a new program in Chapter 5.0, "Administrative Controls," entitled the "Risk-Informed Completion Time Program."
Page 5
 
ATTACHMENT 1 Description and Assessment As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below.
: 1.      Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed changes permit the extension of Completion Times provided the associated risk is assessed and managed in accordance with the NRC approved Risk Informed Completion Time Program. The proposed changes do not involve a significant increase in the probability of an accident previously evaluated because the changes involve no change to the plant or its modes of operation. The proposed changes do not increase the consequences of an accident because the design-basis mitigation function of the affected systems is not changed and the consequences of an accident during the extended Completion Time are no different from those during the existing Completion Time.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2.      Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The proposed 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).
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3.      Does the proposed change involve a significant reduction in a margin of safety?
Response: No The proposed changes permit the extension of Completion Times provided that risk is assessed and managed in accordance with the NRC approved Risk-Informed Completion Time Program. The proposed changes implement a risk-informed configuration management program to assure that adequate margins of safety are maintained. Application of these new specifications and the configuration management program considers cumulative effects of multiple systems or components being out of service and does so more effectively than the current TS.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Page 6
 
ATTACHMENT 1 Description and Assessment Based on the above, EGC concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
3.2      Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's 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.
 
==4.0      ENVIRONMENTAL CONSIDERATION==
 
The proposed changes would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed changes do not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed changes.
Page 7
 
ATTACHMENT 2 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Proposed Technical Specification Changes - Mark-Ups TS Pages 1.0-24          3.3-46          3.6-6        3.7-13 3.1-20          3.3-49          3.6-10          3.8-1 3.3-1          3.3-50          3.6-11          3.8-2 3.3-25          3.3-65          3.6-13          3.8-3 3.3-26          3.3-66          3.6-14        3.8-24 3.3-32          3.3-69          3.6-22        3.8-34 3.3-33          3.3-70          3.6-24        3.8-39 3.3-34          3.3-74          3.6-32        3.8-40 3.3-35          3.3-78          3.6-34        5.0-16c 3.3-36            3.5-1          3.6-59 3.3-37            3.5-2          3.6-62 3.3-45          3.5-11          3.7-1
 
Completion Times 1.3 1.3  Completion Times EXAMPLES          EXAMPLE 1.3-7  (continued)
Condition B is entered, but continues from the time Condition A was initially entered. If Required Action A.1 is met after Condition B is entered, Condition B is exited INSERT 1        and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expired.
IMMEDIATE        When "Immediately" is used as a Completion Time, the COMPLETION TIME  Required Action should be pursued without delay and in a controlled manner.
CLINTON                              1.0-24                  Amendment No.
 
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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
------------------------------------NOTES-------------------------------------
: 1. Separate Condition entry is allowed for each Function.
: 2. When Functions 2.b and 2.c channels are inoperable due to the calculated power exceeding the APRM output by more than 2% RTP while operating at 21.6% RTP, entry into associated Conditions and Required Actions may be delayed for up to 2 hours.
CONDITION                    REQUIRED ACTION        COMPLETION TIME A. One or more Functions  A.1    Place one channel in  48 hours with one channel              affected Function in        RICT INSERT inoperable.                    trip.
B. One or more Functions  B.1    Place one channel in  6 hours with two channels              affected Function in inoperable.                    trip.                        RICT INSERT C. One or more Functions  C.1    Restore two channels  1 hour with three or more            in affected Function channels inoperable.          to OPERABLE status.
D. Required Action and    D.1    Enter the Condition    Immediately associated Completion          referenced in Time of Condition A,          Table 3.3.1.1-1 for B, or C not met.              the channel.
(continued)
CLINTON                            3.3-1                    Amendment No. 222
 
EOC-RPT Instrumentation 3.3.4.1 3.3  INSTRUMENTATION 3.3.4.1  End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation LCO  3.3.4.1    Four channels for each EOC-RPT instrumentation Function listed below shall be OPERABLE:
: a. Turbine Stop Valve (TSV) Closure; and
: b. Turbine Control Valve (TCV) Fast Closure, Trip Oil Pressure-Low.
APPLICABILITY:    THERMAL POWER t 33.3% RTP with any recirculation pump in fast speed.
ACTIONS
------------------------------------NOTE--------------------------------------
Separate Condition entry is allowed for each Function.
CONDITION                  REQUIRED ACTION          COMPLETION TIME A. One or more Functions  A.1  Restore channel to        48 hours with one required            OPERABLE status.
channel inoperable.                                            RICT INSERT OR A.2    --------NOTE---------
Not applicable if inoperable channel is the result of an inoperable breaker.
Place one channel in      48 hours affected Function in RICT INSERT trip.
(continued)
CLINTON                          3.3-25                      Amendment No. 149
 
EOC-RPT Instrumentation 3.3.4.1 ACTIONS  (continued)
CONDITION                  REQUIRED ACTION          COMPLETION TIME B. One or more Functions    B.1    Place one channel in    6 hours with two channels                affected Function in inoperable.                      trip.                          RICT INSERT C. One or more Functions    C.1    Restore two channels    2 hours with three or more              in affected Function channels inoperable.            to OPERABLE status.
D. Required Action and      D.1    Remove the associated    8 hours associated Completion            recirculation pump Time not met.                    fast speed breaker from service.
OR D.2    Reduce THERMAL POWER    8 hours to < 33.3% RTP.
SURVEILLANCE REQUIREMENTS
------------------------------------NOTE--------------------------------------
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains EOC-RPT trip capability.
SURVEILLANCE                              FREQUENCY SR  3.3.4.1.1      Perform CHANNEL FUNCTIONAL TEST.            In accordance with the Surveillance Frequency Control Program (continued)
CLINTON                          3.3-26                      Amendment No. 192
 
ECCS Instrumentation 3.3.5.1 ACTIONS  (continued)
CONDITION                REQUIRED ACTION          COMPLETION TIME B. As required by      B.1  --------NOTE---------
Required Action A.1      Only applicable for and referenced in        Functions 1.a, 1.b, Table 3.3.5.1-1.          2.a, and 2.b.
Declare supported          1 hour from feature(s) inoperable      discovery of when its redundant          loss of feature ECCS initiation    initiation capability is              capability for inoperable.                feature(s) in both divisions AND B.2  --------NOTE---------
Only applicable for Functions 3.a and 3.b.
Declare High Pressure      1 hour from Core Spray (HPCS)          discovery of System inoperable.          loss of HPCS initiation capability AND B.3  Place channel in trip.      24 hours (continued)
RICT/NOTE 1 INSERT CLINTON                        3.3-32                        Amendment No. 216
 
ECCS Instrumentation 3.3.5.1 ACTIONS  (continued)
CONDITION                REQUIRED ACTION        COMPLETION TIME C. As required by      C.1  ---------NOTE--------
Required Action A.1      Only applicable for and referenced in        Functions 1.c, 1.d, Table 3.3.5.1-1.          2.c, and 2.d.
Declare supported        1 hour from feature(s) inoperable    discovery of when its redundant      loss of feature ECCS            initiation initiation capability    capability for is inoperable.          feature(s) in both divisions AND C.2  Restore channel to      24 hours OPERABLE status.
RICT/NOTE 1 INSERT (continued)
CLINTON                        3.3-33                    Amendment No. 216
 
ECCS Instrumentation 3.3.5.1 ACTIONS  (continued)
CONDITION                REQUIRED ACTION      COMPLETION TIME D. As required by      D.1  --------NOTE--------
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    1 hour from inoperable.            discovery of loss of HPCS initiation capability AND D.2.1 Place channel in        24 hour trip.
RICT/NOTE 1 OR                            INSERT D.2.2 Align the HPCS pump    24 hours suction to the suppression pool.
(continued)
CLINTON                        3.3-34                    Amendment No. 193
 
ECCS Instrumentation 3.3.5.1 ACTIONS  (continued)
CONDITION                REQUIRED ACTION          COMPLETION TIME E. As required by      E.1  --------NOTE---------
Required Action A.1      Only applicable for and referenced in        Functions 1.e, 1.f, Table 3.3.5.1-1.          and 2.e.
Declare supported          1 hour from feature(s) inoperable      discovery of when its redundant        loss of feature ECCS              initiation initiation capability      capability for is inoperable.            feature(s) in both divisions AND E.2  Restore channel to        7 days OPERABLE status.
(continued)
RICT/NOTE 1 INSERT CLINTON                        3.3-35                      Amendment No. 216
 
ECCS Instrumentation 3.3.5.1 ACTIONS  (continued)
CONDITION                REQUIRED ACTION                          COMPLETION TIME F. As required by      F.1  Declare Automatic                        1 hour from Required Action A.1      Depressurization                          discovery of and referenced in        System (ADS) valves                      loss of ADS Table 3.3.5.1-1.          inoperable.                              initiation capability in both trip or in accordance with the            systems AND Risk Informed Completion Time Program F.2  Place channel in                          96 hours from trip.                                    discovery of inoperable channel concurrent with HPCS or reactor core isolation cooling (RCIC) inoperable AND 8 days (continued)
                                              --------------NOTE--------------
The Risk Informed Completion Time Program is not applicable when trip capability is not maintained.
8 days or in accordance with the Risk Informed Completion Time Program CLINTON                        3.3-36                                        Amendment No. 95
 
ECCS Instrumentation 3.3.5.1 ACTIONS  (continued)
CONDITION                REQUIRED ACTION                        COMPLETION TIME G. As required by      G.1  --------NOTES--------
Required Action A.1      Only applicable for and referenced in        Functions 4.c, 4.e, Table 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 AND      or in accordance with the Risk Informed Completion Time Program G.2  Restore channel to                      96 hours from OPERABLE status.                        discovery of inoperable channel concurrent with HPCS or RCIC inoperable AND 8 days H. Required Action and  H.1  Declare associated                      Immediately associated Completion      supported feature(s)
Time of Condition B,        inoperable.
C, D, E, F, or G not met.
                                                  --------------NOTE--------------
The Risk Informed Completion Time Program is not applicable when trip capability is not maintained.
8 days or in accordance with the Risk Informed Completion Time Program CLINTON                        3.3-37                                      Amendment No. 95
 
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        B.1  Declare RCIC System      1 hour from Required Action A.1        inoperable.              discovery of and referenced in                                    loss of RCIC Table 3.3.5.3-1.                                      initiation capability AND B.2  Place channel in          24 hours trip.
RICT/NOTE 1 INSERT C. As required by        C.1  Restore channel to        24 hours Required Action A.1        OPERABLE status.
and referenced in Table 3.3.5.3-1.
(continued)
CLINTON                          3.3-45                      Amendment No. 216
 
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 and referenced in          RCIC pump suction is Table 3.3.5.3-1.          not aligned to the suppression pool.
Declare RCIC System      1 hour from inoperable.              discovery of loss of RCIC initiation capability AND D.2.1 Place channel in          24 hours trip.
RICT/NOTE 1 OR                              INSERT D.2.2 Align RCIC pump          24 hours suction to the suppression pool.
E. Required Action and  E.1  Declare RCIC System      Immediately associated Completion      inoperable.
Time of Condition B, C, or D not met.
CLINTON                        3.3-46                      Amendment No. 216
 
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
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each channel.
CONDITION                    REQUIRED ACTION        COMPLETION TIME
-----------NOTE----------    A.1    Place one channel in  48 hours Only applicable to Main              affected Function in RICT INSERT Steam Line (MSL) isolation            trip.
Functions.
A. One or more Functions with one channel inoperable.
-----------NOTE----------    B.1    Place one channel in  6 hours Only applicable to MSL                affected Function in isolation Functions.                  trip.                        RICT INSERT B. One or more Functions with two channels inoperable.
-----------NOTE----------    C.1    Restore two channels  1 hour Only applicable to MSL                in affected Function isolation Functions.                  to OPERABLE status.
C. One or more Functions with three or more channels inoperable.
(continued)
CLINTON                            3.3-49                    Amendment No. 216
 
Primary Containment and Drywell Isolation Instrumentation 3.3.6.1 ACTIONS  (continued)
CONDITION                  REQUIRED ACTION          COMPLETION TIME
------------NOTE----------  D.1    Place channel in trip. 24 hours Not applicable to MSL isolation Functions.                                              RICT/NOTE 1
--------------------------                                        INSERT D. One or more required channels inoperable.
------------NOTE---------    E.1    Restore isolation      1 hour Not applicable to MSL                capability.
isolation Functions.
E. One or more automatic Functions with isolation capability not maintained.
F. Required Action and      F.1    Enter the Condition    Immediately associated Completion            referenced in Time of Condition A,            Table 3.3.6.1-1 for B, C, D, or E not                the channel.
met.
G. As required by          G.1    Isolate associated      12 hours Required Action F.1              MSL.
and referenced in Table 3.3.6.1-1.        OR G.2.1  Be in MODE 3.          12 hours AND G.2.2  Be in MODE 4.          36 hours H. As required by          H.1    Be in MODE 2.          6 hours Required Action F.1 and referenced in Table 3.3.6.1-1.
I. As required by          I.1    Isolate the affected    1 hour Required Action F.1              penetration flow and referenced in                path(s).
Table 3.3.6.1-1.
(continued)
CLINTON                          3.3-50                      Amendment No. 216
 
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 channels    A.1  Enter the Condition      Immediately inoperable.                  referenced in Table 3.3.6.3-1 for the channel.
B. As required by          B.1  Declare associated        1 hour from Required Action A.1          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 B.2  Place channel in          24 hours trip.
RICT/NOTE 1 INSERT (continued)
CLINTON                          3.3-65                        Amendment No. 95
 
RHR Containment Spray System Instrumentation 3.3.6.3 ACTIONS  (continued)
CONDITION                REQUIRED ACTION        COMPLETION TIME C. As required by        C.1  --------NOTE---------
Required Action A.1        Only applicable for and referenced in          Functions 4 and 5.
Table 3.3.6.3-1.            ---------------------
Declare associated      1 hour from RHR containment spray    discovery of subsystem inoperable. loss of RHR containment spray initiation capability in both trip systems AND C.2  Restore channel to        24 hours OPERABLE status.
RICT/NOTE 1 INSERT D. Required Action and  D.1    Declare associated      Immediately associated Completion        RHR containment spray Time of Condition B          subsystem inoperable.
or C not met.
CLINTON                        3.3-66                      Amendment No. 95
 
SPMU System Instrumentation 3.3.6.4 3.3  INSTRUMENTATION 3.3.6.4  Suppression Pool Makeup (SPMU) System Instrumentation LCO  3.3.6.4    The SPMU System instrumentation for each Function in Table 3.3.6.4-1 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 channels    A.1  Enter the Condition        Immediately inoperable.                  referenced in Table 3.3.6.4-1 for the channel.
B. As required by          B.1    Declare associated        1 hour from Required Action A.1            SPMU subsystem            discovery of and referenced in              inoperable.              loss of SPMU Table 3.3.6.4-1.                                        initiation capability in both trip systems AND B.2    Place channel in          24 hours trip.                          RICT/NOTE 1 INSERT (continued)
CLINTON                          3.3-69                        Amendment No. 95
 
SPMU System Instrumentation 3.3.6.4 ACTIONS  (continued)
CONDITION                REQUIRED ACTION        COMPLETION TIME C. As required by        C.1  --------NOTE---------
Required Action A.1        Only applicable for and referenced in          Function 4.
Table 3.3.6.4-1.            ---------------------
Declare associated        1 hour from SPMU subsystem            discovery of inoperable.              loss of SPMU initiation capability in both trip systems AND C.2  Restore channel to        24 hours OPERABLE status.              RICT/NOTE 1 INSERT D. Required Action and  D.1  Declare associated        Immediately associated Completion      SPMU subsystem Time of Condition B        inoperable.
or C not met.
CLINTON                        3.3-70                      Amendment No. 95
 
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.1  Restore trip system        7 days inoperable.                  to OPERABLE status.              RICT INSERT OR A.2  Declare associated        7 days relief and LLS valve(s) inoperable.
B. Required Action and    B.1  Be in MODE 3.              12 hours associated Completion Time of Condition A    AND not met.
B.2  Be in MODE 4.              36 hours OR Two trip systems inoperable.
CLINTON                          3.3-74                      Amendment No. 192
 
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.1    Place channel in trip. 1 hour inoperable.
AND                                    RICT/NOTE 1 INSERT A.2    ---------NOTE----------
Only applicable for Functions 1.c, 1.d, 1.e, 2.c, 2.d, and 2.e
                                    -----------------------    RICT/NOTE 1 INSERT Restore channel to OPERABLE status.          7 days B. Required Action and    B.1    Declare associated DG    Immediately associated Completion          inoperable.
Time not met.
CLINTON                          3.3-78                      Amendment No. 122
 
ECCS  Operating 3.5.1 3.5  EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (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 seven safety/
relief valves shall be OPERABLE.
                  ----------------------------NOTE----------------------------
One low pressure coolant injection (LPCI) subsystem may be inoperable during alignment and operation for decay heat removal with reactor steam dome pressure less than the residual heat removal cut in permissive pressure.
APPLICABILITY:    MODE 1, MODES 2 and 3, except ADS valves are not required to be OPERABLE with reactor steam dome pressure d 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                    ECCS injection/spray        RICT INSERT subsystem inoperable.              subsystem to OPERABLE status.
B. High Pressure Core        B.1      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.
RICT INSERT (continued)
CLINTON                                3.5-1                  Amendment No. 216
 
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            RICT INSERT subsystem to OPERABLE OR                              status.
One ECCS injection and one ECCS spray subsystem inoperable.
D. Required Action and    -------------NOTE------------
associated Completion  LCO 3.0.4.a is not Time of Condition A,  applicable when entering B, or C not met.      MODE 3.
D.1      Be in MODE 3.        12 hours E. One ADS valve          E.1      Restore ADS valve to  14 days inoperable.                    OPERABLE status.            RICT INSERT F. One ADS valve          F.1      Restore ADS valve to  72 hours inoperable.                    OPERABLE status.
RICT INSERT AND                    OR One low pressure ECCS  F.2      Restore low pressure  72 hours injection/spray                ECCS injection/spray RICT INSERT subsystem inoperable.          subsystem to OPERABLE status.
(continued)
CLINTON                            3.5-2                  Amendment No. 187
 
RCIC System 3.5.3 3.5  EMERGENCY CORE COOLING SYSTEMS (ECCS), REACTOR PRESSURE VESSEL (RPV)
WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC)
SYSTEM 3.5.3  RCIC System LCO  3.5.3        The RCIC System shall be OPERABLE.
APPLICABILITY:    MODE 1, MODES 2 and 3 with reactor steam dome pressure > 150 psig.
ACTIONS
-------------------------------------NOTE-------------------------------------
LCO 3.0.4.b is not applicable to RCIC.
CONDITION                  REQUIRED ACTION        COMPLETION TIME A. RCIC System              A.1      Verify by            1 hour inoperable.                        administrative means High Pressure Core Spray System is OPERABLE.
AND A.2      Restore RCIC System  14 days to OPERABLE status.          RICT INSERT B. Required Action and      B.1      Be in MODE 3.        12 hours associated Completion Time not met.            AND B.2      Reduce reactor steam  36 hours dome pressure to d 150 psig.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                            FREQUENCY SR  3.5.3.1    Verify the RCIC System locations              In accordance susceptible to gas accumulation are          with the sufficiently filled with water.              Surveillance Frequency Control Program (continued)
CLINTON                              3.5-11                  Amendment No. 216
 
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Division 1 and 2 SX Subsystems and UHS 3.7.1 3.7  PLANT SYSTEMS 3.7.1  Division 1 and 2 Shutdown Service Water (SX) Subsystems and Ultimate Heat Sink (UHS)
LCO  3.7.1        Division 1 and 2 SX subsystems and the UHS shall be OPERABLE.
APPLICABILITY:    MODES 1, 2, and 3.
ACTIONS CONDITION                  REQUIRED ACTION          COMPLETION TIME A. UHS water volume not    A.1    Restore UHS water        90 days within limit                    volume to within limit.
-----------NOTE-----------  ------------NOTES------------
Not applicable during    1. Enter applicable replacement of                Conditions and Required Division 2 SX pump            Actions of LCO 3.8.1, during the Division 2        "AC Sources -Operating,"
SX system outage              for diesel generator made window from October 26        inoperable by SX.
through November 8, 2015.                    2. Enter applicable
--------------------------        Conditions and Required Actions of LCO 3.4.9, B. Division 1 or 2 SX            "Residual Heat Removal subsystem inoperable.        (RHR) Shutdown Cooling System - Hot Shutdown,"
for RHR shutdown cooling subsystem made inoperable by SX.
B.1    Restore SX subsystem    72 hours to OPERABLE status.            RICT INSERT (continued)
CLINTON                              3.7-1                  Amendment No. 207
 
Main Turbine Bypass 3.7.6 3.7  PLANT SYSTEMS 3.7.6  Main Turbine Bypass System LCO  3.7.6        The Main Turbine Bypass System shall be OPERABLE.
OR The following limits are made applicable:
: a. Reactor THERMAL POWER limit for an inoperable Main Turbine Bypass System as specified in the COLR; and
: b. LCS 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," limit for an inoperable Main Turbine Bypass System as specified in the COLR; and
: c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," limit for an inoperable Main Turbine Bypass System as specified in the COLR.
APPLICABILITY:    THERMAL POWER t 21.6% RTP.
ACTIONS CONDITION                  REQUIRED ACTION        COMPLETION TIME A. Requirements of the      A.1      Satisfy the            2 hours LCO not met.                      requirements of the          RICT INSERT LCO.
B. Required Action and      B.1      Reduce THERMAL POWER  4 hours associated Completion            to < 21.6% RTP.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                            FREQUENCY SR  3.7.6.1    Verify one complete cycle of each main        In accordance turbine bypass valve.                          with the Surveillance Frequency Control Program (continued)
CLINTON                              3.7-13                  Amendment No. 195
 
AC Sources  Operating 3.8.1 3.8  ELECTRICAL POWER SYSTEMS 3.8.1  AC Sources  Operating LCO  3.8.1      The following AC electrical power sources shall be OPERABLE:
: a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electric Power Distribution System; and
: b. Three diesel generators (DGs).
APPLICABILITY:    MODES 1, 2, and 3.
                  ----------------------------NOTE----------------------------
Division 3 AC electrical power sources are not required to be OPERABLE when High Pressure Core Spray System is inoperable.
ACTIONS
-------------------------------------NOTE-------------------------------------
LCO 3.0.4.b is not applicable to DGs.
CONDITION                  REQUIRED ACTION          COMPLETION TIME A. One offsite circuit      A.1      Perform SR 3.8.1.1      1 hour inoperable.                        for OPERABLE offsite circuit.                AND Once per 8 hours thereafter AND A.2      Restore offsite        72 hours circuit to OPERABLE            RICT INSERT status.
(continued)
CLINTON                                  3.8-1                Amendment No. 227
 
AC Sources  Operating 3.8.1 ACTIONS (continued)
CONDITION        REQUIRED ACTION                COMPLETION TIME B. One required DG B.1    Perform SR 3.8.1.1            1 hour inoperable.            for OPERABLE offsite circuit(s).                  AND Once per 8 hours thereafter AND B.2    Declare required              4 hours from feature(s), supported        discovery of by the inoperable DG,        Condition B inoperable when the          concurrent with redundant required            inoperability of feature(s) are                redundant inoperable.                  required feature(s)
AND B.3.1  Determine OPERABLE            24 hours DG(s) are not inoperable due to common cause failure.
OR B.3.2  Perform SR 3.8.1.2            24 hours for OPERABLE DG(s).
AND B.4    Restore required DG          72 hours from to OPERABLE status.          discovery of an inoperable Division 3 DG AND 14 days (continued) or in accordance with the Risk Informed Completion Time Program CLINTON                        3.8-2                      Amendment No. 227
 
AC Sources  Operating 3.8.1 ACTIONS (continued)
CONDITION                REQUIRED ACTION          COMPLETION TIME C. Two offsite circuits  C.1    Declare required        12 hours from inoperable.                    feature(s) inoperable  discovery of when the redundant      Condition C required feature(s)    concurrent with are inoperable.        inoperability of redundant required feature(s)
AND C.2    Restore one offsite    24 hours circuit to OPERABLE          RICT INSERT status.
D. One offsite circuit    D.1    Restore offsite        12 hours inoperable.                    circuit to OPERABLE            RICT INSERT status.
AND OR One required DG inoperable.            D.2    Restore required DG    12 hours to OPERABLE status.            RICT INSERT E. Two required DGs      E.1    Restore one required    2 hours inoperable.                    DG to OPERABLE status.                OR 24 hours if Division 3 DG is inoperable F. Required Action and    -------------NOTE------------
Associated Completion  LCO 3.0.4.a is not applicable Time of Condition A,  when entering MODE 3.
B, C, D, or E not met. -----------------------------
F.1    Be in MODE 3.          12 hours (continued)
CLINTON                              3.8-3                Amendment No. 187
 
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, Division 3, and Division 4 DC electrical power subsystems shall be OPERABLE.
APPLICABILITY:    MODES 1, 2, and 3.
ACTIONS CONDITION                  REQUIRED ACTION          COMPLETION TIME A. One battery charger on    A.1    Restore battery        2 hours Division 1 or 2                  terminal voltage to inoperable.                      greater than or equal to the minimum established float voltage.
AND A.2    Verify battery float    Once per 12 current < 2 amps.      hours AND A.3    Restore battery        7 days charger to OPERABLE RICT INSERT status.
B. One battery on                                            2 hours B.1    Restore battery to Division 1 or 2                  OPERABLE status.              RICT INSERT inoperable.
2 hours C. Division 1 or 2 DC        C.1    Restore Division 1 electrical power                                                RICT INSERT and 2 DC electrical subsystem inoperable            power subsystems to for reasons other                OPERABLE status.
than Condition A or B.
(continued)
CLINTON                              3.8-24                    Amendment No. 187
 
Inverters-Operating 3.8.7 3.8  ELECTRICAL POWER SYSTEMS 3.8.7  Inverters  Operating LCO  3.8.7        The Division 1, 2, 3, and 4 inverters, and A and B RPS solenoid bus inverters shall be OPERABLE.
APPLICABILITY:    MODES 1, 2, and 3.
ACTIONS
-------------------------------------NOTE-------------------------------------
Enter applicable Conditions and Required Actions of LCO 3.8.9, "Distribution Systems  Operating," with any uninterruptible AC bus de-energized.
CONDITION                  REQUIRED ACTION          COMPLETION TIME A. Division 1 or 2          A.1    Restore Division 1      7 days inverter inoperable.            and 2 inverters to            RICT INSERT OPERABLE status.
B. Required Action and      -------------NOTE------------
associated Completion    LCO 3.0.4.a is not Time of Condition A      Applicable when entering not met.                MODE 3.
B.1      Be in MODE 3.          12 hours C. One or more Division 3  C.1    Declare High Pressure    Immediately or 4 inverters                  Core Spray System inoperable.                      inoperable.
D. One RPS solenoid bus    D.1.1    Transfer RPS bus to    1 hour inverter inoperable.              alternate power source.
AND D.1.2    Verify RPS bus supply  Once per 8 hours frequency t 57 Hz.      thereafter OR D.2      De-energize RPS bus. 1 hour (continued)
CLINTON                              3.8-34                    Amendment No. 187
 
Distribution Systems-Operating 3.8.9 3.8  ELECTRICAL POWER SYSTEMS 3.8.9  Distribution Systems  Operating LCO  3.8.9      Division 1, 2, and 3 AC, Division 1, 2, 3, and 4 DC, and Division 1, 2, 3, and 4 uninterruptible AC bus electrical power distribution subsystems shall be OPERABLE.
APPLICABILITY:    MODES 1, 2, and 3.
                  ----------------------------NOTE----------------------------
Division 3 and 4 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      8 hours or 2 AC electrical                and 2 AC electrical          RICT/NOTE 2 power distribution                power distribution            INSERT subsystems inoperable.            subsystems to OPERABLE status.
B. One or more Division 1    B.1      Restore Division 1      8 hours or 2 uninterruptible              and 2 uninterruptible          RICT/NOTE 2 AC bus distribution                AC bus distribution subsystems inoperable.            subsystems to                  INSERT OPERABLE status.
(continued)
CLINTON                              3.8-39                    Amendment No. 227
 
Distribution Systems  Operating 3.8.9 ACTIONS (continued)
CONDITION                REQUIRED ACTION          COMPLETION TIME C. One or more Division 1 C.1      Restore Division 1    2 hours or 2 DC electrical              and 2 DC electrical          RICT/NOTE 2 power distribution              power distribution subsystems inoperable.          subsystems to                INSERT OPERABLE status.
D. Required Action and    -------------NOTE------------
associated Completion  LCO 3.0.4.a is not Time of Condition A,  applicable when entering B, or C not met.      MODE 3.
D.1      Be in MODE 3.          12 hours E. One or more Division 3 E.1      Declare High Pressure  Immediately or 4 AC, DC, or                Core Spray System uninterruptible AC bus          inoperable.
electrical power distribution subsystems inoperable.
F. Two or more divisions  F.1      Enter LCO 3.0.3.      Immediately with inoperable distribution subsystems that result in a loss of function.
CLINTON                            3.8-40                  Amendment No. 227
 
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Clinton Power Station TS Inserts INSERT 1 EXAMPLE  1.3-8 ACTIONS CONDITION          REQUIRED ACTION            COMPLETION TIME A. One subsystem    A.1 Restore subsystem      7 days inoperable.        to OPERABLE status.
OR In accordance with the Risk Informed Completion Time Program B. Required        B.1 Be in MODE 3.          6 hours Action and associated      AND Completion Time not met. B.2 Be in MODE 5.          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. 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.
The Risk Informed Completion Time Program requires recalculation of the RICT to reflect changing plant conditions. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.
For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours after the plant configuration change, whichever is less.
If the 7 day Completion Time clock of Condition A has expired and subsequent changes in plant condition result in exiting the applicability of the Risk Informed Completion Time Program without restoring the inoperable subsystem to OPERABLE status, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start.
If the RICT expires or is recalculated to be less than the elapsed time since the Condition was entered and the inoperable subsystem has not been restored to OPERABLE status, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable subsystems are restored to OPERABLE status after Condition B is entered, Condition A is exited, and therefore, the Required Actions of Condition B may be terminated.
 
Clinton Power Station TS Inserts RICT INSERT OR In accordance with the Risk Informed Completion Time Program RICT/NOTE 1 INSERT OR
      --------NOTE--------
Not applicable when trip capability is not maintained.
In accordance with the Risk Informed Completion Time Program RICT/NOTE 2 INSERT OR
      --------NOTE--------
Not applicable if loss of function.
In accordance with the Risk Informed Completion Time Program RICT/NOTE 3 INSERT OR
      --------NOTE--------
Not applicable if leakage exceeds limits or if loss of function.
In accordance with the Risk Informed Completion Time Program
 
Clinton Power Station TS Inserts INSERT 2 5.5.17  Risk Informed Completion Time Program This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines." The program shall include the following:
: a. The RICT may not exceed 30 days;
: b. A RICT may only be utilized in MODE 1 and 2;
: c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.
: 1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.
: 2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours after the plant configuration change, whichever is less.
: 3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.
: d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:
: 1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or
: 2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.
: e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods used to support this license amendment, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.
 
ATTACHMENT 3 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Proposed Technical Specification Bases Changes - Mark-Ups (For Information Only)
TS Bases Pages B 3.1-39a              B 3.3-163              B 3.6-12            B 3.8-5 B 3.3-20              B 3.3-164              B 3.6-19            B 3.8-9 B 3.3-70              B 3.3-192              B 3.6-20            B 3.8-10 B 3.3-110              B 3.3-193              B 3.6-21            B 3.8-10b B 3.3-112              B 3.3-203              B 3.6-36            B 3.8-11 B 3.3-113              B 3.3-204              B 3.6-40            B 3.8-53 B 3.3-115              B 3.3-211              B 3.6-57            B 3.8-54 B 3.3-116              B 3.3-227              B 3.6-63            B 3.8-71 B 3.3-117              B 3.5-5                B 3.6-111          B 3.8-81 B 3.3-118              B 3.5-6                B 3.6-118          B 3.8-82 B 3.3-130              B 3.5-7                B 3.7-4            B 3.8-84 B 3.3-131a            B 3.5-31                B 3.7-33
 
SLC System B 3.1.7 BASES APPLICABLE          suppression pool pH at or above 7 following a LOCA to SAFETY ANALYSIS      ensure that iodine will be retained in the suppression (continued)        pool water (Ref. 8).
ACTIONS        A.1                      RICT BASES INSERT 1 If one SLC subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this condition, the remaining OPERABLE subsystem is adequate to perform the shutdown function. However, the overall reliability is reduced because a single failure in the (continued)
CLINTON                          B 3.1-39a                  Revision No. 10-5
 
RPS Instrumentation B 3.3.1.1 BASES ACTIONS          GAF for any channel is defined as the power value determined (continued)    by the heat balance divided by the APRM reading for that channel. Upon completion of the gain adjustment, or expiration of the allowed time, the channel must be returned to OPERABLE status or the applicable Condition entered and the Required Actions taken. This Note is based on the time required to perform gain adjustments on multiple channels.
A.1 Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 48 hours has been shown to be acceptable (Ref. 9) to permit restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided the associated Function's inoperable channel is the only inoperable channel and the Function still maintains RPS trip capability (refer to Required Actions B.1 and C.1 Bases.) If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate two failures, 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 a full scram), Condition D must be entered and its Required Action taken.
B.1                                    RICT BASES INSERT 2 Condition B exists when, for any one or more Functions, two required channels are inoperable. In this condition, the RPS still maintains trip capability for that Function, but cannot accommodate a single failure in that Function.
Required Action B.1 limits the time the RPS scram logic for any Function would not accommodate single failure. Within the 6 hour allowance, the associated Function will have at least one inoperable channel in trip.
Completing this Required Action restores RPS to an equivalent reliability level as that evaluated in Reference 9, which justified a 48 hour allowable out of service time.
The 6 hour Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram.
(continued)
RICT BASES INSERT 2 CLINTON                        B 3.3-20                      Revision No. 20-6
 
EOC-RPT Instrumentation B 3.3.4.1 BASES_________________________________________________________________________
ACTIONS          A.1 and A.2  (continued) and allow operation to continue. As noted in Required Action A.2, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable breaker, since this may not adequately compensate for the inoperable breaker (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), Condition D must be entered and its Required Actions taken.
B.1                                      RICT BASES INSERT 2 Condition B exists when, for any one or more Functions, two required channels are inoperable. In this condition, the EOC-RPT still maintains trip capability for that Function, but cannot accommodate a single failure in that Function.
Required Action B.1 limits the time the EOC-RPT logic for any Function would not accommodate single failure. Within the 6 hour allowance, the associated Function will have at least one inoperable channel in trip.
Completing this Required Action restorers EOC-RPT to an equivalent reliability level as that evaluated in Reference 6, which justified a 48 hour allowable out of service time.
The 6 hour Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a recirculation pump trip.
Placing one of the two inoperable channels in trip satisfies both Required Actions A.2 and B.1 for that Function. If one channel is already in trip for the Function when a second channel is determined to be inoperable, Required Action B.1 is met by the one channel already in trip for that function and no additional action is required.
(continued)
RICT BASES INSERT 2 CLINTON                        B 3.3-70                      Revision No. 1-1
 
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS B.1, B.2, and B.3  (continued)
(the Note to Required Action B.1 and Required Action B.2) to delineate which Required Action is applicable for each Function that requires entry into Condition B if an associated channel is inoperable. This ensures that the proper loss of initiation capability check is 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 B.1, the Completion Time only begins upon discovery that a redundant feature in both Divisions (e.g., any Division 1 ECCS and Division 2 ECCS) cannot be automatically initiated due to inoperable, untripped channels within the same variable as described in the paragraph above. For Required Action B.2, the Completion Time only begins upon discovery that the HPCS System cannot be automatically initiated due to two inoperable, untripped channels for the associated Function 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. RICT BASES INSERT 3 Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.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.1 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 (continued)
CLINTON              B 3.3-110                    Revision No. 20-2
 
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS          C.1 and C.2  (continued) if a channel in this Function is inoperable), since the loss of one channel results in a loss of the Function (two-out-of-two logic). This loss was considered during the development of Reference 4 and considered acceptable for the 24 hours allowed by Required Action C.2.
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
For Required Action C.1, the Completion Time only begins upon discovery that the same feature in both Divisions (e.g., any Division 1 ECCS and Division 2 ECCS) cannot be automatically initiated due to inoperable channels within the same variable as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.
Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would either cause the initiation or would not necessarily result in a safe state for the channel in all events.
D.1, D.2.1, and D.2.2                RICT BASES INSERT 3 Required Action D.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 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 D.2.1 and D.2.2 is not appropriate and the HPCS System must be declared inoperable within 1 hour after 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.
(continued)
CLINTON                        B 3.3-112                      Revision No. 1-1
 
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS          D.1, D.2.1, and D.2.2  (continued)
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
For Required Action D.1, the Completion Time only begins upon discovery that the 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.
RICT BASES INSERT 3 Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1 or the suction source must be aligned to the suppression pool per 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 D.2.1 or Required Action D.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 D.2.1 and D.2.2, Condition H must be entered and its Required Action taken.
within 24 hours E.1 and E.2 Required Action E.1 is intended to 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.1, the features would be those that are initiated by Functions 1.e, 1.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, 1.f, (continued)
CLINTON                        B 3.3-113                        Revision No. 0
 
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS          E.1 and E.2  (continued)
If the instrumentation that controls the pump minimum flow valve is inoperable such that the valve will not automatically open, pump operation with no injection path available could lead to pump overheating and failure. If there were a failure of the instrumentation  such that the valve would not automatically close, a portion of the pump flow could be diverted from the reactor injection path, causing insufficient core cooling. 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 DBA 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.
RICT BASES INSERT 3 F.1 and F.2 Required Action F.1 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) more than one Function 4.a channel and one Function 5.a channel are inoperable and untripped, (b) one Function 4.b channel 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 (continued)
CLINTON                        B 3.3-115                    Revision No. 1-1
 
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS            F.1 and F.2  (continued)
                    "time zero" for beginning the allowed outage time "clock."
For Required Action F.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable, untripped channels within similar ADS trip system Functions as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
RICT BASES INSERT 2 Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable channel to OPERABLE status if both HPCS and RCIC are OPERABLE. If either HPCS or RCIC is inoperable, the time is RICT BASES INSERT 3 shortened to 96 hours. If the status of HPCS or RCIC changes such that the Completion Time changes from 8 days to 96 hours, the 96 hours begins upon discovery of HPCS or RCIC inoperability. However, total time for an inoperable, 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.1 and G.2 Required Action G.1 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 (continued)
CLINTON                          B 3.3-116                        Revision No. 0
 
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS          G.1 and G.2  (continued) 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 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.1 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.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable channels within similar ADS trip system Functions, as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be acceptable (Ref. 4) to permit restoration of any inoperable channel to OPERABLE status if both HPCS and RCIC are OPERABLE (Required Action G.2). If either HPCS or RCIC is (continued)
RICT BASES INSERT 3 CLINTON                        B 3.3-117                      Revision No. 5-9
 
ECCS Instrumentation B 3.3.5.1 BASES                                RICT BASES INSERT 2 ACTIONS          G.1 and G.2  (continued) inoperable, the time is reduced to 96 hours. If the status of HCPS or RCIC changes such that the Completion Time changes from 8 days to 96 hours, the 96 hours begins upon discovery of HPCS or RCIC inoperability. However, total time for an inoperable 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 channel. 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.
H.1 With any Required Action and associated Completion Time not met, the associated feature(s) may be incapable of performing the intended function and the supported feature(s) associated with the inoperable untripped channels must be declared inoperable immediately.
SURVEILLANCE      As noted at the beginning of the SRs, the SRs for each ECCS REQUIREMENTS      instrumentation Function are found in the SRs column of Table 3.3.5.1-1.
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 Conditions and Required Actions may be delayed for up to 6 hours as follows: (a) for Functions 3.c, 3.f, 3.g, and 3.h; and (b) for Functions other than 3.c, 3.f, 3.g, and 3.h provided the associated Function or the redundant Function maintains ECCS initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions (continued)
CLINTON                        B 3.3-118                        Revision No. 0
 
RCIC System Instrumentation B 3.3.5.3 BASES ACTIONS B.1 and B.2 (continued)
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.
RICT BASES INSERT 3 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. 2) 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.
C.1 A risk based analysis was performed and determined that an allowable out of service time of 24 hours (Ref. 2) is acceptable to permit restoration of any inoperable channel to OPERABLE status (Required Action C.1). A Required Action (similar to Required Action B.1), limiting the allowable out of service time if a loss of automatic RCIC initiation capability exists, is not required. This Condition applies to the Reactor Vessel Water Level-High, Level 8 Function, whose logic is arranged such that any inoperable channel will result in a loss of automatic RCIC initiation capability. As stated above, this loss of automatic RCIC initiation capability was analyzed and determined to be acceptable. This Condition also applies to the Manual Initiation Function. Since this Function is not assumed in any accident or transient analysis, a total loss of manual initiation capability (Required Action C.1) for 24 hours is (continued)
CLINTON              B 3.3-130                    Revision No. 20-2
 
RCIC System Instrumentation B 3.3.5.3 BASES ACTIONS            D.1, D.2.1, and D.2.2 (continued)
(continued)                                          RICT BASES INSERT 3 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. 2) 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 within 24 hours channel (shifting the suction source to the suppression pool). 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 Conditions and Required Actions may be delayed as follows:
(continued)
CLINTON                        B 3.3-131a                      Revision No. 20-2
 
Primary Containment and Drywell Isolation Instrumentation B 3.3.6.1 BASES ACTIONS          A.1  (continued)
RICT BASES INSERT 2 inoperable channel and the Function still maintains isolation capability (refer to Required Action B.1 and C.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate two failures, 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 undesired isolation), Condition F must be entered and its Required Action taken.
B.1 Condition B exists when, for any one or more MSL isolation Functions, two required channels are inoperable. (For example, a failure of a coincidence logic card (i.e., a two-out-of-four logic card) in one division may affect two channels). In this condition, the MSL isolation system still maintains isolation capability for that Function, but cannot accommodate an additional single failure in that Function.
Required Action B.1 limits the time the MSL isolation logic for any Function would not accommodate a single failure.
Within the 6 hour allowance, the associated Function will have at least one inoperable channel in trip. Completing this Required Action restores the MSL isolation system to an equivalent reliability level as that evaluated in Reference 6, which justified a 48 hour allowable out of service time.
The 6 hour Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of MSL isolation. Placing one of the two inoperable channels in trip satisfies both Required Actions A.1 and B.1 for that Function. If one channel is already in trip for the Function when a second channel is determined to be inoperable, Required Action B.1 is met by the one channel already in trip for that Function and no additional action is required.
RICT BASES INSERT 2 Alternately, if it is not desired to place one inoperable channel in trip (e.g., as in the case where placing the (continued)
CLINTON                          B 3.3-163                      Revision No. 9-8
 
Primary Containment and Drywell Isolation Instrumentation B 3.3.6.1 BASES ACTIONS          B.1  (continued) inoperable channel in trip would result in an undesired isolation, scram, or RPT), Condition F must be entered and its Required Action taken.
C.1 Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels for the same Function result in the Function not being able to accommodate a single failure and maintain MSL isolation capability. A Function is considered to be maintaining MSL isolation capability when sufficient channels are OPERABLE or in trip such that the Function will generate a trip signal on a valid signal. For a Function with two-out-of-four logic, this would require the Function to have three channels OPERABLE or in trip.
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
D.1                          RICT BASES INSERT 3 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 24 hours has been shown to be acceptable (Refs. 5 and 6) to permit restoration of any inoperable channel to OPERABLE status for any Function other than MSL isolation. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action E.1 Bases). 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.1. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue with no further restrictions. 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 isolation),
Condition F must be entered and its Required Action taken.
(continued)
CLINTON                          B 3.3-164                        Revision No. 0
 
RHR Containment Spray System Instrumentation B 3.3.6.3 BASES ACTIONS          B.1 and B.2  (continued) 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 of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
RICT BASES INSERT 3 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 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 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.1 and C.2 Required Action C.1 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 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 initiation capability for both trip systems. As noted, Required Action C.1 is only applicable for Functions 2 and 4. The Required Action is not applicable to Function 5 or 6 (which also requires entry into this Condition if a channel in one of these Functions is inoperable). Function 5 does not have a corresponding channel in both subsystems. Function 6 is the Manual Initiation Function and is not assumed in any USAR accident (continued)
CLINTON                          B 3.3-192                        Revision No. 0
 
RHR Containment Spray System Instrumentation B 3.3.6.3 BASES ACTIONS          C.1 and C.2  (continued) or transient analysis. Thus, a total loss of manual initiation capability for 24 hours (as allowed by Required Action C.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 C.1, 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.
RICT BASES INSERT 3 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 permit restoration of any inoperable channel to OPERABLE status.
If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition D must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action could either cause the initiation or it would not necessarily result in a safe state for the channel in all events.
D.1 With any Required Action and associated Completion Time not met, the associated RHR containment spray subsystem may be incapable of performing the intended function and the RHR containment spray subsystem associated with inoperable untripped channels must be declared inoperable immediately.
SURVEILLANCE      As noted at the beginning of the SRs, the SRs for each RHR REQUIREMENTS      Containment Spray System Function are located in the SRs column of Table 3.3.6.3-1.
The Surveillances are also modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into (continued)
CLINTON                        B 3.3-193                          Revision No. 0
 
SPMU System Instrumentation B 3.3.6.4 BASES ACTIONS          B.1 and B.2  (continued) not appropriate and both SPMU subsystems must be declared inoperable within 1 hour after discovery of loss of SPMU 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 B.1, the Completion Time only begins upon discovery that the SPMU 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 of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
RICT BASES INSERT 3 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 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 D must be entered and its Required Action taken.
C.1 and C.2 Required Action C.1 is intended to ensure appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic initiation capability for the SPMU System. In this case, automatic initiation capability is lost if 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 the SPMU System must be declared inoperable within 1 hour after discovery of loss of SPMU initiation capability for both (continued)
CLINTON                          B 3.3-203                      Revision No. 0
 
SPMU System Instrumentation B 3.3.6.4 BASES ACTIONS          C.1 and C.2  (continued) trip systems. As noted, Required Action C.1 is only applicable for Function 4. Required Action C.1 is not applicable to Function 5 (which also requires entry into this Condition if a channel in this Function is inoperable),
since it is the Manual Initiation Function and is not assumed in any USAR accident or transient analysis. Thus, a total loss of manual initiation capability for 24 hours (as allowed by Required Action C.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 C.1, the Completion Time only begins upon discovery that the SPMU 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.
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 permit restoration of any inoperable channel to OPERABLE status.
If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition D must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action could either cause the initiation or it would not necessarily result in a safe state for the channel in all events.
RICT BASES INSERT 3 D.1 With any Required Action and associated Completion Time not met, the associated SPMU subsystem may be incapable of performing the intended function and the SPMU subsystem associated with inoperable, untripped channels must be declared inoperable immediately.
(continued)
CLINTON                        B 3.3-204                        Revision No. 0
 
Relief and LLS Instrumentation B 3.3.6.5 BASES LCO              of MSIV closure with a direct scram (i.e., MSIV position (continued)    switches). This analysis is also described in Reference 1.
APPLICABILITY    The relief and LLS instrumentation is required to be OPERABLE in MODES 1, 2, and 3, since considerable energy exists in the nuclear steam system and the S/RVs may be needed to provide pressure relief. If the S/RVs are needed, then the relief and LLS functions are required to ensure that the primary containment design basis is maintained. In MODES 4 and 5, the reactor pressure is low enough that the overpressure limit cannot be approached by assumed operational transients or accidents. Thus, pressure relief and LLS instrumentation are not required.
ACTIONS          A.1 and A.2 Because the failure of any reactor steam dome pressure instrument channels in one trip system will not prevent the associated S/RV from performing its relief and LLS functions, 7 days is allowed to restore a trip system to OPERABLE status (Required Action A.1). In this condition, the remaining OPERABLE trip system is adequate to perform the relief and LLS initiation functions. 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.
RICT BASES INSERT 2 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 functions 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.
(continued)
CLINTON                        B 3.3-211                        Revision No. 0
 
LOP Instrumentation B 3.3.8.1 BASES ACTIONS          A.1 and A.2    RICT BASES INSERT 3 (continued)
With one or more channels of a Function inoperable, the Function may not be capable of performing the intended function. Therefore, only 1 hour is allowed to restore the inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. For Loss of Voltage Functions, placing the inoperable channel in trip would conservatively compensate for the inoperability and allow operation to continue. However, for Degraded Voltage Functions, placing the inoperable channel in trip may not conservatively compensate for the inoperability. Because of the assumptions used in the setpoint calculations, the setpoint(s) for the remaining OPERABLE channel(s) may not ensure reset of the relay within the required voltage range.
As a result, operation with an inoperable Degraded Voltage channel(s) in trip is limited to 7 days.
RICT BASES INSERT 4 Thus, 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) or if the inoperable channel(s) is not restored to OPERABLE status within the allowable out of service time, 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 Completion Times are acceptable because they minimize risk while allowing time for restoration or tripping of channels.
Required Action A.2 is modified by a Note which states that the Required Action is only applicable for Functions 1.c, 1.d, 1.e, 2.c, 2.d, and 2.e. The 7-day limitation is imposed as a result of assumptions associated with the setpoint calculations for the modified Degraded Voltage Function instrumentation.
(continued)
CLINTON                        B 3.3-227                    Revision No. 4-3
 
ECCS !Operating B 3.5.1 BASES LCO            required core cooling, thereby allowing operation of RHR (continued) shutdown cooling when necessary.
APPLICABILITY  All ECCS subsystems are required to be OPERABLE during MODES 1, 2, and 3 when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system piping. In MODES 2 and 3, the ADS function is not required when pressure is  150 psig because the low pressure ECCS subsystems (LPCS and LPCI) are capable of providing flow into the RPV below this pressure. ECCS requirements for MODES 4 and 5 are specified in LCO 3.5.2, "RPV Water Inventory Control."
ACTIONS        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.
A.1                                RICT BASES INSERT 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).
B.1 and B.2                RICT BASES INSERT 1 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 (continued)
CLINTON                          B 3.5-5                  Revision No. 20-2
 
ECCS !Operating B 3.5.1 BASES ACTIONS B.1 and B.2  (continued) 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.
C.1                                    RICT BASES INSERT 1 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.
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 the overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 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 prohibits the application of LCO 3.0.4.a. This Note clarifies the intent of the Required Action by indicating that it is not (continued)
CLINTON                  B 3.5-6                  Revision No. 20-2
 
ECCS !Operating B 3.5.1 BASES ACTIONS D.1  (continued) permissible under LCO 3.0.4.a to enter MODE 3 from MODE 4 with the LCO not met. While remaining in MODE 3 presents an acceptable level of risk, it is not the intent of the Required Action to allow entry into, and continue operation in, MODE 3 from MODE 4 in accordance with LCO 3.0.4.a.
However, where allowed, a risk assessment may be performed in accordance with LCO 3.0.4.b. Consideration of the results of this risk assessment is required to determine the acceptability of entering MODE 3 from MODE 4 when this LCO is not met. 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.
E.1 The LCO requires seven 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 six 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.
RICT BASES INSERT 2 F.1 and F.2 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.
RICT BASES INSERT 2 (continued)
CLINTON                  B 3.5-7                    Revision No. 20-2
 
RCIC System B 3.5.3 BASES ACTIONS        A.1 and A.2                              RICT BASES INSERT 1 (continued)
If the RCIC System is inoperable during MODE 1, or MODES 2 or 3 with reactor steam dome pressure > 150 psig, and the HPCS System is verified to be OPERABLE, the RCIC System must be restored to OPERABLE status within 14 days. In this Condition, 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 loss of coolant accident (LOCA). OPERABILITY of the HPCS is therefore verified within 1 hour when the RCIC System is inoperable. This may be performed as an administrative check, by examining logs or other information, to determine if the HPCS is out of service for maintenance or other reasons. Verification does not require performing the Surveillances needed to demonstrate the OPERABILITY of the HPCS System. If the OPERABILITY of the HPCS System cannot be verified, however, Condition B must be immediately entered. For transients and certain abnormal events with no LOCA, RCIC (as opposed to HPCS) is the preferred source of makeup coolant because of its relatively small capacity, which allows easier control of RPV water level. Therefore, a limited time is allowed to restore the inoperable RCIC to OPERABLE status.
The 14 day Completion Time is based on a reliability study (Ref. 3) that evaluated the impact on ECCS availability, 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 (AOTs). Because of the similar functions of the HPCS and RCIC, the AOTs (i.e., Completion Times) determined for the HPCS are also applied to RCIC.
B.1 and B.2 If the RCIC System cannot be restored to OPERABLE status within the associated Completion Time, or if the HPCS System is simultaneously inoperable, the plant must be brought to a condition 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 reactor steam dome pressure reduced to 150 psig within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
(continued)
CLINTON                        B 3.5-31                  Revision No. 20-2
 
Primary Containment Air Locks B 3.6.1.2 BASES ACTIONS C.1, C.2, and C.3  (continued) 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 OPERABLE status within 1 hour.
Additionally, the air lock must be restored to OPERABLE status within 24 hours. 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.
RICT BASES INSERT 5 D.1 and D.2 If the inoperable primary containment air lock cannot be restored to OPERABLE status within the associated Completion Time while operating 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.
E.1 and E.2 If the inoperable primary containment air lock cannot be restored to OPERABLE status within the associated Completion Time during movement of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours) in the primary or secondary containment, action is required to immediately suspend activities that represent a potential for releasing radioactive material, thus placing the unit in a Condition that minimizes risk. If applicable, movement of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours) must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.
The Required Actions of Condition E are modified by a Note indicating that LCO 3.0.3 does not apply. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations.
Therefore, inability to suspend movement of recently irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.
(continued)
CLINTON                    B 3.6-12                Revision No. 20-2
 
PCIVs B 3.6.1.3 BASES                                                RICT BASES INSERT 2 ACTIONS A.1 and A.2  (continued)        RICT BASES INSERT 1 the valve secured. For penetrations isolated in accordance with Required Action A.1, the device used to isolate the penetration should be the closest one available to the primary containment. The Required Action must be completed within the 4 hour Completion Time (8 hours for main steam lines and 12 hours for instrument line excess flow check valves (EFCVs)). 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 for plant shutdown. For EFCVs, a 12 hour Completion Time is allowed. The Completion Time of 12 hours for EFCVs allows a period of time to restore the EFCVs to OPERABLE status given the fact that these valves are associated with instrument lines which are of small diameter and thus represent less significant leakage paths.
RICT BASES INSERT 2 For affected penetrations that have been isolated in accordance with Required Action A.1, 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 the correct position. The Completion Time for this verification of "once per 31 days for 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.            following isolation 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.
(continued)
CLINTON                      B 3.6-19                  Revision No. 20-2
 
PCIVs B 3.6.1.3 BASES ACTIONS        B.1 (continued)
With one or more penetration flow paths with two PCIVs inoperable, except due to leakage not within limits, 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 affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange.
The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1.1.
C.1 With the secondary containment bypass leakage rate, 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 restored 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.1, 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.1, 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 (refer to the requirements of SR 3.6.1.3.1; if this requirement is not met, entry into Condition A and B, as appropriate, would also be required), so that a gross breach of primary containment does not exist.
RICT BASES INSERT 2 In accordance with Required Action D.2, this penetration flow path must be verified to be isolated on a periodic following isolation (continued)
CLINTON                              B 3.6-20            Revision No. 20-2
 
PCIVs B 3.6.1.3 BASES ACTIONS D.1, D.2, and D.3  (continued) basis. The periodic verification is necessary to ensure that 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 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.      following isolation For a primary containment purge valve with a 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. The normal Frequency for SR 3.6.1.3.5 is as required by the Primary Containment Leakage Rate Testing Program. 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.
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.
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 (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours) in the primary and secondary containment must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe condition.
(continued)
CLINTON                    B 3.6-21              Revision No. 20-2
 
LLS Valves B 3.6.1.6 BASES APPLICABLE      LLS S/RVs are specified, all five LLS S/RVs do not operate SAFETY ANALYSES in any DBA analysis.LLS valves satisfy Criterion 3 of the (continued) NRC Policy Statement LCO            Five LLS valves are required to be OPERABLE to satisfy the assumptions of the safety analysis (Ref. 1). The requirements of this LCO are applicable to the mechanical and electrical/pneumatic capability of the LLS valves to function for controlling the opening and closing of the S/RVs.
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.
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.
RICT BASES INSERT 2 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 the overall plant risk is minimized. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 2) 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 B.1 is modified by a Note that prohibits the application of LCO 3.0.4.a. This Note clarifies the intent of the Required Action by indicating that it is not permissible under LCO 3.0.4.a to enter MODE 3 from MODE 4 with the LCO not met. While remaining in MODE 3 presents an acceptable level of risk, it is not the intent of the Required Action to allow entry into, and continue operation in, MODE 3 from MODE 4 in accordance with LCO 3.0.4.a.
(continued)
CLINTON                              B 3.6-36              Revision No. 13-2
 
RHR Containment Spray System B 3.6.1.7 BASES APPLICABLE      The analysis demonstrates that with containment spray SAFETY ANALYSES operation the containment pressure remains within design (continued)  limits.
The RHR Containment Spray System satisfies Criterion 3 of the NRC Policy Statement.
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. Management of gas voids is important to RHR Containment Spray System OPERABILITY.
The LCO is modified by a Note that allows an RHR containment spray subsystem to be inoperable during alignment and operation for decay heat removal with reactor steam dome pressure less than the residual heat removal cut-in permissive pressure. This is necessary since the RHR system is required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor, and manual realignment from the shutdown cooling mode to the RHR containment spray mode could result in pump cavitation and voiding in the suction piping, resulting in the potential to damage the RHR system, including water hammer. One RHR Containment Spray subsystem is allowed to be considered inoperable for this temporary period, because in shutdown cooling mode it is fulfilling a decay heat removal capacity function. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling when necessary.
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.
ACTIONS        A.1                RICT BASES INSERT 1 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 (continued)
CLINTON                            B 3.6-40                Revision No. 20-1
 
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, heat exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE.
Management of gas voids is important to RHR Suppression Pool Cooling System OPERABILITY.
The LCO is modified by a Note that allows one RHR suppression pool cooling subsystem to be inoperable during alignment and operation for decay heat removal with reactor steam dome pressure less than the residual heat removal cut-in permissive pressure. This is necessary since the RHR system is required to operate in the shutdown cooling mode to remove decay heat and sensible heat from the reactor, and manual realignment from the shutdown cooling mode to the RHR suppression pool cooling mode could result in pump cavitation and voiding in the suction piping, resulting in the potential to damage the RHR system, including water hammer. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling for decay heat removal.
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                RICT BASES INSERT 1 With one RHR suppression pool cooling subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining 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.
(continued)
CLINTON                            B 3.6-57                Revision No. 20-1
 
SPMU System B 3.6.2.4 BASES ACTIONS        B.1 (continued)
When upper containment pool water temperature is > 120qF, the heat absorption capacity is inadequate to ensure that the suppression pool heat sink capability matches the safety analysis assumptions. Increased temperature has a relatively smaller impact on heat sink capability.
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                                    RICT BASES INSERT 1 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.
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)
CLINTON                            B 3.6-63              Revision No. 14-2
 
Drywell Air Lock B 3.6.5.2 BASES ACTIONS B.1, B.2, and B.3  (continued)
The Required Actions are modified by two Notes. Note 1 ensures only the Required Actions and associated Completion Times of Condition C are required if both doors in the air lock are inoperable. Note 2 allows entry and exit into the drywell under the control of a dedicated individual stationed at the air lock to ensure that only one door is opened at a time (i.e., the individual performs the function of the interlock). In addition, Note 2 allows an OPERABLE air lock door to remain unlocked, but closed, when the door is under the control of a dedicated individual stationed at the air lock.
C.1 and C.2 With the air lock inoperable for reasons other than those described in Condition A or B, Required Action C.1 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.
RICT BASES INSERT 5 Additionally, the air lock must be restored to OPERABLE status within 24 hours. 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 the air lock.
(continued)
CLINTON                    B 3.6-111                Revision No. 20-2
 
Drywell Isolation Valves B 3.6.5.3 BASES ACTIONS A.1 and A.2  (continued)
The associated system piping is adequate to perform the isolation function for other drywell penetrations. However, the overall reliability is reduced because a single failure 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. 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.
RICT BASES INSERT 2 For affected penetration flow paths that have been isolated in accordance with Required Action A.1, 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 valve 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 the Completion Time specified for PCIVs in LCO 3.6.1.3, Primary Containment Isolation Valves (PCIVs).
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 controls.
Allowing verification by administrative controls 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.
(continued)
CLINTON                    B 3.6-118              Revision No. 20-2
 
Division 1 and 2 SX Subsystems and UHS B 3.7.1 BASES (continued)
ACTIONS          A.1 If the UHS is inoperable (i.e., the UHS water volume is not within the limit), action must be taken to restore the inoperable UHS to OPERABLE status within 90 days. The 90 day Completion Time is reasonable considering the time required to restore the required UHS volume, the margin contained in the available heat removal capacity, and the low probability of a DBA occurring during this period.
B.1                  RICT BASES INSERT 1 If the Division 1 or 2 SX subsystem is inoperable, it must be restored to OPERABLE status within 72 hours. With the unit in this condition, the remaining OPERABLE Division 1 or 2 SX subsystem is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure in the OPERABLE Division 1 or 2 SX subsystem could result in loss of the SX 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.
Condition B is modified by a Note. The Note indicates that this Condition is not applicable during replacement of the Division 2 SX pump during the Division 2 SX system outage window from October 26 through November 8, 2015.
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 SX 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.
C.1 During replacement of the Division 2 SX pump in the Division 2 SX system outage window from October 26 through November 8, 2015, the Division 2 SX subsystem is inoperable, and it must be restored to OPERABLE status within 7 days.
This Completion Time is based upon a risk-informed assessment that concluded that the associated risk with the system in the specified configuration is acceptable.
Condition C is modified by a Note. The Note indicates that this Condition is only applicable during replacement of the (continued)
CLINTON                            B 3.7-4                    Revision No. 17-2
 
Main Turbine Bypass System B 3.7.6 BASES LCO          is not exceeded. With an inoperable Main Turbine Bypass (continued) System, reactor power may be limited in accordance with the cycle-dependent COLR and modification to MCPR limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") AND LHGR limits (LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)") may be applied in accordance with the cycle-dependent COLR to allow continued operation. An OPERABLE Main Turbine Bypass System requires the bypass valves to open in response to increasing main steam line pressure. This response is within the assumptions of the applicable analysis (Ref. 2).
The reactor power limitations and the MCPR and LHGR limits for an inoperable Main Turbine Bypass System are specified in the COLR.
APPLICABILITY The Main Turbine Bypass System is required to be OPERABLE at
              ! 21.6% RTP to ensure that the fuel cladding integrity Safety Limit and the cladding 1% plastic strain limit are not violated during the feedwater controller failure, maximum demand event. As discussed in the Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," sufficient margin to these limits exists
              < 21.6% RTP. Therefore, these requirements are only necessary when operating at or above this power level.
ACTIONS      A.1 If the Main Turbine Bypass System is inoperable (one or more bypass valves inoperable), and the reactor power limit for an inoperable Main Turbine Bypass System, and the MCPR and LHGR limits for an inoperable Main Turbine Bypass System, as specified in the COLR, are not applied, the assumptions of the design basis transient analysis may not be met. Under such circumstances, prompt action should be taken to restore the Main Turbine Bypass System to OPERABLE status or limit reactor power and apply the MCPR and LHGR limits as specified in the COLR. The 2 hour Completion Time is reasonable, based on the time to complete the Required Action and the low probability of an event occurring during this period requiring the Main Turbine Bypass System.
B.1 RICT BASES INSERT 2 If the Main Turbine Bypass System cannot be restored to OPERABLE status within the associated Completion Time, or the reactor power limit for an inoperable Main Turbine Bypass System, as specified in the COLR is not applied, and the MCPR and LHGH limits for an inoperable Main Turbine (continued)
CLINTON                        B 3.7-33                  Revision No. 14-3
 
AC Sources  Operating B 3.8.1 BASES APPLICABILITY A Note has been added taking exception to the Applicability (continued) requirements for Division 3 sources, provided the HPCS System is declared inoperable. This exception is intended to allow declaring of the HPCS System inoperable either in lieu of declaring the Division 3 source inoperable, or at any time subsequent to entering ACTIONS for an inoperable Division 3 source. This exception is acceptable since, with the HPCS System inoperable and the associated ACTIONS entered, the Division 3 AC sources provide no additional assurance of meeting the above criteria.
AC power requirements for MODES 4 and 5 are covered in LCO 3.8.2, "AC Sources  Shutdown."
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.
A.1 To ensure a highly reliable power source remains, it is necessary to verify the availability of the remaining 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 circuit fails SR 3.8.1.1, the second offsite circuit is inoperable, and Condition C, for two offsite circuits inoperable, is entered.
A.2              RICT BASES INSERT 2 According to Regulatory Guide 1.93 (Ref. 6), operation may continue in Condition A for a period that should not exceed 72 hours. With one offsite circuit inoperable, the reliability of the offsite system is degraded, and the potential for a loss of offsite power is increased, with attendant potential for a challenge to the plant safety systems. In this Condition, however, the remaining OPERABLE offsite circuit and DGs are adequate to supply electrical power to the onsite Class 1E distribution system.
The Completion Time takes into account the capacity and capability of the remaining AC sources, reasonable time for repairs, and the low probability of a DBA occurring during this period.
(continued)
CLINTON                        B 3.8-5                  Revision No. 10-3
 
AC Sources  Operating B 3.8.1 BASES ACTIONS          B.4 (continued)
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. RICT BASES INSERT 2 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 the low probability of a DBA 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 for an inoperable Division 3 DG when Condition B may have already been entered for another equipment inoperability and is still in effect).
The second Completion Time (14 days)  applies to an inoperable Division 1 or 2 DG and is  a risk-informed allowed out-of-service time (AOT) based on a  plant-specific risk analysis performed to establish this  AOT for the Division 1 and 2 DGs.
The evaluation that supports this Completion Time considered both planned and unplanned DG outage time. Based on this evaluation, it is intended that use of the full, 14-day completion time would be limited to once per DG per cycle (24 months) to perform a planned DG overhaul.
To mitigate increased risk during the period beyond 72 hours and up to 14 days, the following actions must be completed prior to exceeding 72 hours:
x Verification that the RAT and ERAT are operable.
x Verification of the correct breakers alignment and indicated power availability for each offsite circuit.
x The DG extended Completion Time will not be entered for scheduled maintenance purposes if severe weather conditions are expected.
x Additional elective equipment maintenance or testing that requires the equipment to be removed from service will be evaluated and activities that yield unacceptable results will be avoided.
x The condition of the offsite power supply and switchyard, including transmission lines and ring bus breakers, will be evaluated.
x No elective maintenance will be scheduled within the switchyard that would challenge the RAT connection or offsite power availability.
(continued)
CLINTON                            B 3.8-9                  Revision No. 10-7
 
AC Sources  Operating B 3.8.1 BASES ACTIONS            B.4    (continued) x  Operating crews will be briefed on the DG work plan with consideration given to actions that would be required in the event of a loss of offsite power or station blackout.
C.1 and C.2 RICT BASES INSERT 2 Required Action C.1 addresses actions to be taken in the event of concurrent failure of redundant required features.
Required Action C.1 reduces the vulnerability to a loss of function. The rationale for the 12 hours is that Regulatory Guide 1.93 (Ref. 6) allows a Completion Time of 24 hours for two required offsite circuits inoperable, based upon the assumption that two complete safety divisions are OPERABLE.
When a concurrent redundant required feature failure exists, this assumption is not the case, and a shorter Completion Time of 12 hours is appropriate. These features are designed with redundant safety related divisions (i.e.,
single division systems are not included in the list, although, for this Required Action, Division 3 is considered redundant to Division 1 and 2 ECCS). Redundant required features failures consist of any of these features that are inoperable, because any inoperability is on a division redundant to a division with inoperable offsite circuits.
(continued)
CLINTON                              B 3.8-10                Revision No. 20-10
 
AC Sources  Operating B 3.8.1 BASES ACTIONS C.1 and C.2  (continued)
: b. The time required to detect and restore an unavailable offsite power source is generally much less than that required to detect and restore an unavailable onsite AC source.
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 DBA 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.
RICT BASES INSERT 2 (continued)
CLINTON                  B 3.8-10b                  Revision No. 3-5
 
AC Sources  Operating B 3.8.1 BASES ACTIONS      D.1 and D.2 (continued)
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 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 DBA occurring during this period.
RICT BASES INSERT 2 E.1 With two DGs inoperable, there is one remaining 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.
(continued)
CLINTON                        B 3.8-11                    Revision No. 0
 
DC Sources  Operating B 3.8.4 BASES ACTIONS    A.1, A.2 and A.3 (continued)
(continued) A discharged battery having terminal voltage of at least the minimum established float voltage indicates that the battery is on the exponential charging current portion (the second part) of its recharge cycle. The time to return a battery to its fully charged state under this condition is simply a function of the amount of the previous discharge and the recharge characteristic of the battery. Thus, there is good assurance of fully recharging the battery within 12 hours, avoiding a premature shutdown with its own attendant risk.
If established battery terminal float voltage cannot be restored to greater than or equal to the minimum established float voltage within 2 hours, and the charger is not operating in the current-limiting mode, a faulty charger is indicated. A faulty charger that is incapable of maintaining established battery terminal float voltage does not provide assurance that it can revert to and operate properly in the current limit mode that is necessary during the recovery period following a battery discharge event that the DC system is designed for.
If the charger is operating in the current limit mode after 2 hours that is an indication that the battery is partially discharged and its capacity margins will be reduced. The time to return the battery to its fully charged conditions in this case is a function of the battery charger capacity, the amount of loads on the associated DC system, the amount of the previous discharge, and the recharge characteristic of the battery. The charge time can be extensive, and there is not adequate assurance that it can be recharged with 12 hours (Required Action A.2).
Required Action A.2 requires that the battery float current be verified as less than or equal to 2 amps. This indicates that, if the battery had been discharged as the result of the inoperable battery charger, it has now been fully recharged. If at the expiration of the initial 12 hour period, the battery float current is not less than or equal to 2 amps, this indicates there may be additional battery problems and the battery must be declared inoperable.
Required Action A.3 limits the restoration time for the inoperable battery charger to 7 days. This action is applicable if an alternate means of restoring battery terminal voltage to greater than or equal to the minimum established float voltage has been used (e.g., balance of plant non-Class 1E battery charger). The output of the swing charger will be capable of being connected to any one of the Class 1E DC buses for Division 1, 2 or 4 using a 400 Amp disconnect switch. The connection through this switch will be provided with padlocks on the switches which will be administratively controlled to allow connection to only on of the Class 1E divisions at any time. The 7 day completion time reflects a reasonable time to effect restoration of the qualified battery charger to operable status.
RICT BASES INSERT 1      (continued)
CLINTON                      B 3.8-53                    Revision No. 7-2
 
DC Sources  Operating B 3.8.4 BASES ACTIONS (continued) B.1 Condition B represents one division with one battery inoperable. With one battery inoperable, the DC bus is being supplied by the OPERABLE battery charger. Any event that results in a loss of the AC buss supporting the battery charger will also result in loss of DC to that division.
Recovery of the AC bus, especially if it is due to a loss of offsite power, will be hampered by the fact that many of the components necessary for the recovery (e.g., diesel generator control and field flash, AC load shed and diesel generator output circuit breakers, etc.) likely rely upon the battery.
In addition, the energization transients of any DC loads that are beyond the capability of the battery charger and normally require the assistance of the battery will not be able to be brought online. The 2 hour limit allows sufficient time to effect restoration of an inoperable battery given that the majority of the conditions that lead to battery inoperability (e.g., loss of battery charger, battery cell voltage less than 2.07 V, etc.) are identified in Specifications 3.8.4, 3.8.5, and 3.8.6 together with additional specific completion times.
RICT BASES INSERT 2 C.1 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.
RICT BASES INSERT 2 If one of the required Division 1 or 2 DC electrical power subsystems is inoperable for reasons other than Condition A or B (e.g., inoperable battery charger and associated inoperable battery), the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition. Since a subsequent worst case single failure could, however, result in the loss of minimum necessary DC electrical subsystems, continued power operation should not exceed 2 hours. The 2 hour Completion Time is based on Regulatory Guide 1.93 (Ref. 7) and reflects a reasonable time to assess unit status as a function of the inoperable DC electrical power subsystem and, if the DC electrical power subsystem is not restored to OPERABLE status, to prepare to effect an orderly and safe unit shutdown.
(continued)
CLINTON                      B 3.8-54                    Revision No. 6-5
 
Inverters  Operating B 3.8.7 BASES APPLICABILITY    Inverter requirements for MODES 4 and 5 are covered in the (continued)    Bases for LCO 3.8.8, "Inverters  Shutdown."
ACTIONS          With a required inverter inoperable, its associated uninterruptible AC bus is inoperable if not energized.
LCO 3.8.9 addresses this action; however, pursuant to LCO 3.0.6, these actions would not be entered even if the uninterruptible AC bus were de-energized. Therefore, the ACTIONS are modified by a Note stating that ACTIONS for LCO 3.8.9 must be entered immediately. This ensures the uninterruptible bus is re-energized within 8 hours.
A.1 RICT BASES INSERT 2 Required Action A.1 allows 7 days to restore an inoperable inverter and return it to service. The 7 day limit is a risk-informed Completion Time based on a plant-specific risk analysis performed to establish this Completion Time for the Division 1 and 2 inverters. This risk has to be balanced against the risk of an immediate shutdown, along with the potential challenges to safety systems that such a shutdown might entail. When the uninterruptible AC bus is powered from its constant voltage source, it is relying upon interruptible AC electrical power sources (offsite and onsite). The uninterruptible inverter source to the uninterruptible AC buses is the preferred source for powering instrumentation trip setpoint devices.
An inverter may be removed from service to perform planned preventive maintenance so long as the inverter is restored to operable status within 24 hours (this is an administrative limit intended to allow preventive maintenance to be performed). The intent of the 7 day limit (i.e., the extended completion time (CT) beyond the initial 24 hours) is to restore an inoperable inverter following an inverter failure (i.e., to support online corrective maintenance).
With a required inverter inoperable, the following compensatory actions will be taken:
1.  Entry into Required Action A.1 will not be planned concurrent with Emergency Diesel Generator (EDG) maintenance on the associated train.
2.  Entry into Required Action A.1 will not be planned concurrent with planned maintenance on another RPS or ECCS/RCIC actuation logic channel that could result in that channel being in a tripped condition.
These actions are taken because it is recognized that with an inverter inoperable and the instrument bus being powered by the regulating transformer, instrument power for that train is dependent on power from the associated EDG following a loss of offsite power event.
(continued)
CLINTON                            B 3.8-71                  Revision No. 13-2
 
Distribution Systems  Operating B 3.8.9 BASES  (continued)
ACTIONS            A.1 With one or more Division 1 or 2 required AC buses, load centers, motor control centers, or distribution panels (except uninterruptible AC buses), 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 hours.
RICT BASES INSERT 6 The Condition A worst scenario is one division without AC power (i.e., no offsite power to the division and the associated DG inoperable). In this Condition, the unit is more vulnerable to a complete loss of AC power. It is, therefore, imperative that the unit operators' attention be focused on minimizing the potential for loss of power to the remaining division by stabilizing the unit, and on restoring power to the affected division. The 8 hour time limit before requiring a unit shutdown in this Condition is acceptable because:
: a. There is potential for decreased safety if the unit operators' attention is diverted from the evaluations and actions necessary to restore power to the affected division to the actions associated with taking the unit to shutdown within this time limit.
: b. The potential for an event in conjunction with a single failure of a redundant component in the division with AC power. (The redundant component is verified OPERABLE in accordance with Specification 5.5.10, "Safety Function Determination Program (SFDP).")
(continued)
CLINTON                              B 3.8-81                  Revision No. 20-10
 
Distribution Systems  Operating B 3.8.9 BASES ACTIONS            B.1 (continued)
With one or more Division 1 or 2 uninterruptible AC bus inoperable, the remaining OPERABLE uninterruptible AC buses are capable of supporting the minimum safety functions necessary to shut down and maintain the unit in the safe shutdown condition. Overall reliability is reduced, however, because an additional single failure could result in the minimum required ESF functions not being supported.
Therefore, the required uninterruptible AC bus distribution subsystems must be restored to OPERABLE status within 8 hours by powering the bus from the associated Class 1E constant voltage source transformer.
RICT BASES INSERT 6 Condition B may represent one uninterruptible AC bus distribution subsystem without power; potentially both the DC source and the associated AC source nonfunctioning. In this situation, the plant is significantly more vulnerable to a complete loss of all noninterruptible 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 uninterruptible AC buses, and restoring power to the affected buses.
This 8 hour limit is more conservative than Completion Times allowed for the majority of components that are without adequate uninterruptible AC power. Taking exception to LCO 3.0.2 for components without adequate uninterruptible AC (continued)
CLINTON                            B 3.8-82                Revision No. 20-10
 
Distribution Systems  Operating B 3.8.9 BASES ACTIONS      C.1 (continued)
With one or more Division 1 or 2 DC electrical power distribution subsystems in one division 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. RICT BASES INSERT 6 Condition C 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; (continued)
CLINTON                        B 3.8-84                Revision No. 20-10
 
Clinton Power Station TS Bases Inserts RICT BASES INSERT 1 or in accordance with the Risk Informed Completion Time Program RICT BASES INSERT 2 Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
RICT BASES INSERT 3 Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. A Note has been provided to indicate that the Risk Informed Completion Time Program is not applicable when trip capability is not maintained.
RICT BASES INSERT 4 or in accordance with the Risk Informed Completion Time Program. A Note has been provided to indicate that the Risk Informed Completion Time Program is not applicable when trip capability is not maintained RICT BASES INSERT 5 or in accordance with the Risk Informed Completion Time Program. A Note has been provided to indicate that the Risk Informed Completion Time Program is not applicable if leakage exceeds limits or if there is a loss of function RICT BASES INSERT 6 or in accordance with the Risk Informed Completion Time Program. A Note has been provided to indicate that the Risk Informed Completion Time Program is not applicable if there is a loss of function
 
ATTACHMENT 4 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications
 
ATTACHMENT 4 Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications Tech Spec Description                TSTF-505      CPS        Apply  Comments TS          TS        RICT?
Completion Times                  1.3        1.3 Example 1.3-8                      Example    Example            N/A    TSTF-505 changes are incorporated.
1.3-8      1.3-8 Standby Liquid Control (SLC)      3.1.7      3.1.7 System One SLC subsystem inoperable.      3.1.7.B.1  3.1.7.A.1          Yes    TSTF-505 changes are incorporated.
Reactor Protection System (RPS)    3.3.1.1    3.3.1.1 Instrumentation One or more required channels      3.3.1.1.A.1 -----              No    The CPS TS do not contain these Conditions. Therefore, inoperable.                        3.3.1.1.A.2 -----              No    a change is not proposed to the CPS TS.
One or more Functions with one or  3.3.1.1.B.1 -----              No    The CPS TS do not contain these Conditions. Therefore, more required channels inoperable  3.3.1.1.B.2 -----              No    a change is not proposed to the CPS TS.
in both trip systems.
One or more Functions with one    -----      3.3.1.1.A.1        Yes    The CPS RPS logic is different than the standard BWR/6 channel inoperable.                                                      design upon which NUREG-1434 is based. For CPS, trip capability for each function is maintained with any two of four RPS channels operable. Additional discussion is provided in Attachment 5.
One or more Functions with two    -----      3.3.1.1.B.1        Yes    The CPS RPS logic is different than the standard BWR/6 channels inoperable.                                                    design upon which NUREG-1434 is based. For CPS, trip capability for each function is maintained with any two of four RPS channels operable. Additional discussion is provided in Attachment 5.
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 excluded. This function is not inoperable in MODE 2 with                                                modeled.
intermediate range monitors (IRMs) on Range 2 or below.
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ATTACHMENT 4 Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications Tech Spec Description                    TSTF-505      CPS        Apply  Comments TS          TS        RICT?
End of Cycle Recirculation Pump        3.3.4.1    3.3.4.1 Trip (EOC-RPT) Instrumentation One or more required channels          3.3.4.1.A.1 -----              No    The CPS TS do not contain these Conditions. Therefore, inoperable.                            3.3.4.1.A.2 -----              No    a change is not proposed to the CPS TS.
One or more Functions with one        -----      3.3.4.1.A.1        Yes  The CPS EOC-RPT logic is different than the standard required channel inoperable.          -----      3.3.4.1.A.2        Yes  BWR/6 design upon which NUREG-1434 is based. For CPS, trip capability for each function is maintained with any two of four EOC-RPT channels operable. Additional discussion is provided in Attachment 5.
One or more Functions with two        -----      3.3.4.1.B.1        Yes  The CPS EOC-RPT logic is different than the standard channels inoperable.                                                        BWR/6 design upon which NUREG-1434 is based. For CPS, trip capability for each function is maintained with any two of four EOC-RPT channels operable. Additional discussion is provided in Attachment 5.
Anticipated Transient Without          3.3.4.2    3.3.4.2 Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation One or more channels inoperable.      3.3.4.2.A.1 -----              No    The CPS TS do not contain these Conditions. Therefore, 3.3.4.2.A.2 -----              No    a change is not proposed to the CPS TS.
Emergency Core Cooling System          3.3.5.1    3.3.5.1 (ECCS) Instrumentation As required by Required Action A.1    3.3.5.1.B.3 3.3.5.1.B.3        Yes  TSTF-505 changes are incorporated.
and referenced in Table 3.3.5.1-1.
(Functions 1.a, 1.b, 2.a, and 2.b; 3.a Under certain circumstances, with more than one channel and 3.b).
inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
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ATTACHMENT 4 Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications Tech Spec Description                  TSTF-505        CPS          Apply  Comments TS            TS          RICT?
As required by Required Action A.1  3.3.5.1.C.2  3.3.5.1.C.2        Yes  TSTF-505 changes are incorporated.
and referenced in Table 3.3.5.1-1.
(Functions 1.c, 1.d, 2.c, and 2.d).
Under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
As required by Required Action A.1  3.3.5.1.D.2.1 3.3.5.1.D.2.1      Yes  TSTF-505 changes are incorporated.
and referenced in Table 3.3.5.1-1.
(Functions 3.d and 3.e).
Under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
As required by Required Action A.1  3.3.5.1.E.2  3.3.5.1.E.2        Yes  TSTF-505 changes are incorporated.
and referenced in Table 3.3.5.1-1.
(Functions 1.e, 1.f, and 2.e).
Under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
As required by Required Action A.1  3.3.5.1.F.2  3.3.5.1.F.2        Yes  TSTF-505 changes are incorporated. RICT insert format and referenced in Table 3.3.5.1-1.                                          is modified from TSTF-505 R2 to align with CPS TS 1.2, (Functions 4.a, 4.b, 4.d, 5.a, 5.b,                                          "Logical Connectors," direction to only use first level logic and 5.d).                                                                    for Completion Time.
Under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
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ATTACHMENT 4 Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications Tech Spec Description                      TSTF-505        CPS          Apply Comments TS            TS          RICT?
As required by Required Action A.1      3.3.5.1.G.2  3.3.5.1.G.2        Yes  TSTF-505 changes are incorporated. RICT insert format and referenced in Table 3.3.5.1-1.                                              is modified from TSTF-505 R2 to align with CPS TS 1.2, (Functions 4.c, 4.e, 4.f, 4.g, 5.c, 5.e,                                        "Logical Connectors," direction to only use first level logic and 5.f).                                                                      for Completion Time.
Under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
Reactor Core Isolation Cooling          3.3.5.2      3.3.5.3 (RCIC) System Instrumentation As required by Required Action A.1      3.3.5.2.B.2  3.3.5.3.B.2        Yes  TSTF-505 changes are incorporated.
and referenced in Table 3.3.5.2-1.
(Function 1).
Under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
As required by Required Action A.1      3.3.5.2.D.2.1 3.3.5.3.D.2.1      Yes  TSTF-505 changes are incorporated.
and referenced in Table 3.3.5.2-1.
(Functions 3 and 4).
Under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
Primary Containment Isolation            3.3.6.1      3.3.6.1 Instrumentation One or more required channels            3.3.6.1.A.1  3.3.6.1.D.1        Yes  TSTF-505 changes are incorporated.
inoperable. (Functions 2.b, 5.c, 5.d, and 5.e; Functions other than Under certain circumstances, with more than one channel Functions 2.b, 5.c, 5.d, and 5.e).
inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
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ATTACHMENT 4 Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications Tech Spec Description                      TSTF-505      CPS          Apply  Comments TS          TS          RICT?
One or more Functions with one          -----      3.3.6.1.A.1        Yes    The CPS MSL isolation logic is different than the standard channel inoperable. (Functions 1.a,                                          BWR/6 design upon which NUREG-1434 is based. For 1.b, 1.c, 1.d, 1.e, 1.f, and 1.g).                                            CPS, trip capability for each function is maintained with any two of four MSL isolation channels operable.
Additional discussion is provided in Attachment 5.
One or more Functions with two          -----      3.3.6.1.B.1        Yes    The CPS MSL isolation logic is different than the standard channels inoperable. (Functions                                              BWR/6 design upon which NUREG-1434 is based. For 1.a, 1.b, 1.c, 1.d, 1.e, 1.f, and 1.g).                                      CPS, trip capability for each function is maintained with any two of four MSL isolation channels operable.
Additional discussion is provided in Attachment 5.
Residual Heat Removal (RHR)            3.3.6.3    3.3.6.3 Containment Spray System Instrumentation As required by Required Action A.1      3.3.6.3.B.2 3.3.6.3.B.2        Yes    TSTF-505 changes are incorporated.
and referenced in Table 3.3.6.3-1.
(Functions 1 and 3).
(Functions 1,            Under certain circumstances, with more than one channel 2, and 3).                inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
As required by Required Action A.1      3.3.6.3.C.2 3.3.6.3.C.2        Yes    TSTF-505 changes are incorporated.
and referenced in Table 3.3.6.3-1.
(Functions 2 and 4).
(Functions 4,            Under certain circumstances, with more than one channel 5, and 6).                inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
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ATTACHMENT 4 Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications Tech Spec Description                TSTF-505      CPS          Apply  Comments TS          TS          RICT?
Suppression Pool Makeup            -----        3.3.6.4 (SPMU) System Instrumentation As required by Required Action A.1 -----        3.3.6.4.B.2        Yes    TSTF-505 does not apply a RICT to TS 3.3.6.4, Required and referenced in Table 3.3.6.4-1.                                        Action B.2. Industry guidance developed by the Technical Specifications Task Force indicates that this action was excluded because the traveler will not modify Required Actions that represent a loss of function.
However, CPS proposes to apply a RICT to this action.
The proposed TS change includes a Note to indicate that the Risk Informed Completion Time Program is not applicable when trip capability is not maintained.
As required by Required Action A.1 -----        3.3.6.4.C.2        Yes    TSTF-505 does not apply a RICT to TS 3.3.6.4, Required and referenced in Table 3.3.6.4-1.                                        Action C.2. Industry guidance developed by the Technical Specifications Task Force indicates that this action was excluded because the traveler will not modify Required Actions that represent a loss of function.
However, CPS proposes to apply a RICT to this action.
The proposed TS change includes a Note to indicate that the Risk Informed Completion Time Program is not applicable when trip capability is not maintained.
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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ATTACHMENT 4 Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications Tech Spec Description                TSTF-505        CPS        Apply  Comments TS            TS        RICT?
Loss of Power (LOP)                3.3.8.1      3.3.8.1 Instrumentation One or more channels inoperable. 3.3.8.1.A.1  3.3.8.1.A.1        Yes    TSTF-505 changes are incorporated.
                                  -----        3.3.8.1.A.2        Yes Under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
Safety/Relief Valves (S/RVs)      3.4.4        3.4.4 One required S/RV inoperable.      3.4.4.A.1    -----              No    The CPS TS do not contain this Condition. Therefore, a change is not proposed to the CPS TS.
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.
injection/spray subsystem inoperable.
High Pressure Core Spray (HPCS)    3.5.1.B.2    3.5.1.B.2          Yes    TSTF-505 changes are incorporated.
System inoperable.
Two ECCS injection subsystems      3.5.1.C.1    3.5.1.C.1          Yes    TSTF-505 changes are incorporated.
inoperable OR one ECCS injection and one ECCS spray subsystem inoperable.
One ADS valve inoperable.          3.5.1.E.1    3.5.1.E.1          Yes    TSTF-505 changes are incorporated.
One ADS valve inoperable AND      3.5.1.F.1    3.5.1.F.1          Yes    TSTF-505 changes are incorporated.
one low pressure ECCS              3.5.1.F.2    3.5.1.F.2          Yes injection/spray subsystem inoperable.
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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ATTACHMENT 4 Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications Tech Spec Description                  TSTF-505          CPS        Apply  Comments TS            TS        RICT?
Primary Containment Air Locks        3.6.1.2      3.6.1.2 One or more primary containment      3.6.1.2.C.3  3.6.1.2.C.3        Yes    TSTF-505 changes are incorporated.
air locks inoperable for reasons other than Condition A or B.
Under certain circumstances, with more than one primary containment airlock inoperable, excessive leakage or a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when leakage exceeds limits or there is a loss of function.
Primary Containment Isolation        3.6.1.3      3.6.1.3 Valves (PCIVs)
One or more penetration flow paths  3.6.1.3.A.1  3.6.1.3.A.1        Yes    TSTF-505 changes are incorporated. RICT insert format with one PCIV inoperable except      3.6.1.3.A.2  3.6.1.3.A.2        --    is modified from TSTF-505 R2 to align with CPS TS 1.2, due to leakage not within limit.                                            "Logical Connectors," direction to only use first level logic for Completion Time.
One or more penetration flow paths  3.6.1.3.C.2  -----              No    The CPS TS do not contain this Condition. Therefore, a with one PCIV inoperable for                                                change is not proposed to the CPS TS.
reasons other than Conditions D and E. (Only applicable to penetration flow paths with only one PCIV).
One or more penetration flow paths  3.6.1.3.E.1  3.6.1.3.D.1        Yes    TSTF-505 changes are incorporated.
with one or more containment purge  3.6.1.3.E.2  3.6.1.3.D.2        --
valves not within purge valve 3.6.1.3.E.3  3.6.1.3.D.3        --
leakage limits.
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ATTACHMENT 4 Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications Tech Spec Description                TSTF-505        CPS          Apply  Comments TS            TS          RICT?
Low-Low Set (LLS) Valves          3.6.1.6      3.6.1.6 One LLS valve inoperable.          -----        3.6.1.6.A.1        Yes    TSTF-505 does not apply a RICT to TS 3.6.1.6, Required Action A.1. Industry guidance developed by the Technical Specifications Task Force indicates that this action was excluded because the traveler will not modify Required Actions for systems that do not affect CDF or LERF or for which a RICT cannot be quantitatively determined.
However, CPS proposes to apply a RICT to this action because the function is modeled in the PRA and can be directly included in the RTR tool for the RICT program.
As stated in the CPS TS Bases for Required Action A.1, with one LLS valve inoperable, the remaining operable LLS valves are adequate to perform the designed function.
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.
subsystem inoperable.
Residual Heat Removal (RHR)        3.6.2.3      3.6.2.3 Suppression Pool Cooling One RHR suppression pool cooling  3.6.2.3.A.1  3.6.2.3.A.1        Yes    TSTF-505 changes are incorporated.
subsystem inoperable.
Suppression Pool Makeup            3.6.2.4      3.6.2.4 (SPMU) System One SPMU subsystem inoperable      3.6.2.4.C.1  3.6.2.4.C.1        Yes    TSTF-505 changes are incorporated.
for reasons other than Condition A or B.
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ATTACHMENT 4 Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications Tech Spec Description                  TSTF-505        CPS        Apply  Comments TS          TS        RICT?
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.
reasons other than Condition A or B.
Drywell Isolation Valves              3.6.5.3    3.6.5.3 One or more penetration flow paths    3.6.5.3.A.1 3.6.5.3.A.1        Yes  TSTF-505 changes are incorporated.
with one drywell isolation valve inoperable.
[Standby Service Water (SSW)]        3.7.1      3.7.1 System and [Ultimate Heat Sink (UHS)]
[One or more cooling towers with      3.7.1.A.1  -----              No    The CPS TS do not contain this Condition. Therefore, a one cooling tower fan inoperable.]                                          change is not proposed to the CPS TS.
One [SSW] subsystem inoperable        3.7.1.C.1  3.7.1.B.1          Yes  TSTF-505 changes are incorporated.
[for reasons other than Condition A].
Main Turbine Bypass System            3.7.6      3.7.6 Requirements of the LCO not met.      3.7.6.A.1  3.7.6.A.1          Yes  TSTF-505 changes are incorporated.
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.
inoperable.
One required DG inoperable.          3.8.1.B.4  3.8.1.B.4          Yes  TSTF-505 changes are incorporated. RICT insert format is modified from TSTF-505 R2 to align with CPS TS 1.2, "Logical Connectors," direction to only use first level logic for Completion Time.
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ATTACHMENT 4 Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications Tech Spec Description                  TSTF-505          CPS        Apply    Comments TS            TS        RICT?
Two required offsite circuits        3.8.1.C.2    3.8.1.C.2          Yes    TSTF-505 changes are incorporated.
inoperable.
One required offsite circuit        3.8.1.D.1    3.8.1.D.1          Yes    TSTF-505 changes are incorporated.
inoperable AND one required DG      3.8.1.D.2    3.8.1.D.2          Yes inoperable.
One required automatic load          3.8.1.F.1    -----              No    The CPS TS do not contain this Condition. Therefore, a sequencer inoperable.                                                        change is not proposed to the CPS TS.
DC Sources - Operating              3.8.4        3.8.4 One battery charger on one division  3.8.4.A.3    3.8.4.A.3          Yes    TSTF-505 changes are incorporated.
inoperable.
One battery on one division          3.8.4.B.1    3.8.4.B.1          Yes    TSTF-505 changes are incorporated.
inoperable.
Division 1 or 2 DC electrical power  3.8.4.C.1    3.8.4.C.1          Yes    TSTF-505 changes are incorporated.
subsystem inoperable for reasons other than Condition A or B.
Inverters - Operating                3.8.7        3.8.7 Division 1 or 2 inverter inoperable. 3.8.7.A.1    3.8.7.A.1          Yes    TSTF-505 changes are incorporated.
Distribution Systems - Operating    3.8.9        3.8.9 One or more Division 1 or 2 AC      3.8.9.A.1    3.8.9.A.1          Yes    TSTF-505 changes are incorporated.
electrical power distribution subsystems inoperable.
Under certain circumstances, with more than one AC electrical power distribution subsystem inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when a loss of function occurs.
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ATTACHMENT 4 Cross-Reference of TSTF-505 and Clinton Power Station Technical Specifications Tech Spec Description                  TSTF-505        CPS        Apply    Comments TS            TS        RICT?
One or more Division 1 and 2 AC    3.8.9.B.1    -----              No    The CPS TS do not contain this Condition. Therefore, a vital buses inoperable.                                                    change is not proposed to the CPS TS.
One or more Division 1 or 2        -----        3.8.9.B.1          Yes    This is a plant-specific Condition with a Required Action to uninterruptible AC bus distribution                                        restore the Division 1 and 2 uninterruptible AC bus subsystems inoperable.                                                      distribution subsystems to operable status. CPS proposes to apply a RICT to the existing CPS TS 3.8.9, Required Action B.1.
This is acceptable because the TSTF states that there may also be plant-specific TS to which changes of the type presented in the TSTF may be applied.
Under certain circumstances, with more than one uninterruptible AC bus distribution subsystem inoperable, loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when a loss of function occurs.
One or more Division 1 and 2 DC    3.8.9.C.1    3.8.9.C.1          Yes    TSTF-505 changes are incorporated.
electrical power distribution subsystems inoperable.
Under certain circumstances, with more than one DC electrical power distribution subsystem inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when a loss of function occurs.
Programs and Manuals                5.5          5.5 Programs and Manuals - Risk        [NEW TS]      [NEW TS]          N/A    The CPS TS do not currently contain this program. The Informed Completion Time Program    5.5.15        5.5.17                    new RICT Program will be added to the CPS TS 5.5.17 consistent with TSTF-505.
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ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity ATTACHMENT 5 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Information Supporting Instrumentation Redundancy and Diversity
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity This Attachment provides the components available to respond to identified accident conditions.
Not all components are implied, assumed or directly credited in the Updated Safety Analysis Report (USAR) Chapter 15 Accident Analysis; however, the components have been confirmed to be available and useable.
Technical Specifications (TS) Section 3.3, Instrumentation, contains specific Limiting Conditions for Operation (LCOs). The following TS Instrumentation LCOs are included in this license amendment request (LAR) for Clinton Power Station (CPS), Unit 1.
: 1.      Reactor Protection System (RPS) Instrumentation - LCO 3.3.1.1
: 2.      End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation - LCO 3.3.4.1
: 3.      Emergency Core Cooling System (ECCS) Instrumentation - LCO 3.3.5.1
: 4.      Reactor Core Isolation Cooling (RCIC) System Instrumentation - LCO 3.3.5.3
: 5.      Primary Containment and Drywell Isolation Instrumentation - LCO 3.3.6.1
: 6.      Residual Heat Removal (RHR) Containment Spray System Instrumentation - LCO 3.3.6.3
: 7.      Suppression Pool Makeup (SPMU) System Instrumentation - LCO 3.3.6.4
: 8.      Relief and Low-Low Set (LLS) Instrumentation - LCO 3.3.6.5
: 9.      Loss-of-Power (LOP) Instrumentation - LCO 3.3.8.1 CPS TS Section 3.3, Instrumentation, LCOs were developed to ensure that CPS maintains necessary redundancy and diversity and complies with the "single failure" design criterion as defined in IEEE-279-1971, and the diversity requirements as defined in 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants" (GDC), GDC-22, "Protection System Independence."
Included below is a description of the redundant and/or diverse means available to mitigate accidents that each identified instrumentation and control function defined in TS Section 3.3 is designed to prevent. A table is provided for each TS Instrumentation Section to identify the USAR transient/accident that credits the instrumentation and control function. For each function, diverse instrumentation is identified to demonstrate that at least one diverse means is available for each risk-informed individual functional unit credited in the safety analysis.
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ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
: 1.      Reactor Protection System (RPS)
 
==Reference:==
TS LCO 3.3.1.1 and TS B3.3.1.1 Reactor Protection System (RPS) Instrumentation The RPS design creates defense-in-depth from the redundancy of the channels for each trip system. The RPS is a fail-safe system design (de-energizes to trip) and is comprised of four independent trip logic divisions (1, 2, 3, and 4) as described in USAR section 7.2. There are four sensor channels for each variable, although more than one sensor per variable may provide inputs to each trip channel. The sensor trip channels are designated as A, B, C and D, or divisions 1, 2, 3, and 4. The sensor trip channels are combined into a two-out-of-four logic using isolation modules to assure that no single failure can prevent the required safety action from the remainder of the system. The outputs of the logic channels in a trip system are combined in a two out of four-logic arrangement so any 2 of the 4 channels being in a tripped condition produce a full reactor SCRAM.
The diverse inputs causing a trip of the RPS are: (USAR Figures 7.2-2 thru 7.2-4, CPS TS Table 3.3.1.1-1 and Bases B3.3.1.1).
* Intermediate Range Monitor Neutron Flux High
* 2 channels per division. 4 groups (8 channels) per Function
* Intermediate Range Monitor - INOP
* 2 channels per division. 4 groups (8 channels) per Function
* Average Power Range Monitor Neutron Flux - High, Setdown
* 1 channel per division. 4 channels per Function
* Average Power Range Monitor Flow Biased Simulated Thermal Power - High
* 1 channel per division. 4 channels per Function
* Average Power Range Monitor Fixed Neutron Flux - High
* 1 channel per division. 4 channels per Function
* Average Power Range Monitor - Inop
* 1 channel per division. 4 channels per Function
* Reactor Vessel Steam Dome Pressure - High
* 1 channel per division. 4 channels per Function
* Reactor Vessel Level - Low, Level 3
* 1 channel per division. 4 channels per Function
* Reactor Vessel Level - High, Level 8
* 1 channel per division. 4 channels per Function
* Main Steam Isolation Valve - Closure
* 2 channels per division. 4 groups (8 channels) per Function
* Drywell Pressure - High
* 1 channel per division. 4 channels per Function Page 2
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
: 1.      Reactor Protection System (RPS) - cont.
* SCRAM Discharge Volume Water Level - High Transmitter
* 1 channel per division. 4 channels per Function Float Switch
* 1 channel per division. 4 channels per Function
* Turbine Stop Valve - Closure
* 1 channel per division. 4 channels per Function
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
* 1 channel per division. 4 channels per Function
* Reactor Mode Switch - Shutdown Position
* 1 channel per division. 4 channels per Function
* Manual SCRAM
* 1 channel per division. 4 channels per Function In addition to RPS de-energizing the scram pilot valves and energizing the backup scram valves, CPS has redundant and diverse methods of shutting down the reactor in the unlikely event that the RPS does not SCRAM the reactor. The Anticipated Transient Without SCRAM (ATWS) Alternate Rod Insertion (ARI) subsystems provide backup capability to insert the control rods into the reactor and can be manually or automatically initiated. The Recirculation Pump Trip (RPT) breakers also trip the reactor recirculation pumps to reduce the reactor power via negative void coefficient reactivity feedback via the ATWS-RPT subsystem. CPS also has a Standby Liquid Control System (SLCS) as an independent backup system. The system can be manually initiated via the Main Control Room keylock switches to inject boron into the Reactor Vessel and to initiate closure of the Reactor Water Clean-Up (RWCU) outboard (1G33F004) or inboard (1G33F001) suction isolation valve to prevent removal of the injected boron.
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ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity RPS Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.1.1 1. Intermediate Range Monitors
: a. Neutron Flux -    15.1.6            Inadvertent RHR        1) Automatic Initiation High                                Shutdown Cooling          - APRM Fixed Neutron Flux - High Operation                  - APRM Neutron Flux - High, Setdown
                                                                            - IRM Neutron Flux - High
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown
: b. Inop                None              None                    1) Manual SCRAM
: 2. Average Power Range Monitor
: a. Neutron Flux -    15.1.6            Inadvertent RHR        1) Automatic Initiation High, Setdown                        Shutdown Cooling          - APRM Fixed Neutron Flux - High Operation                  - APRM Neutron Flux High, Setdown
                                                                            - IRM Neutron Flux - High
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown 15.4.9            Control Rod Drop        1) Automatic Initiation Accident                  - APRM Fixed Neutron Flux - High
                                                                            - APRM Neutron Flux - High, Setdown Trip
* IRM Neutron Flux - High
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown
: b. Flow Biased        15.1.1            Loss of Feedwater      1) Automatic Initiation Simulated                            Heating                    - APRM Flow Biased Simulated Thermal Power - Upscale Thermal
* APRM Fixed Neutron Flux - High Power - Upscale                                                - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown Page 4
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity RPS Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                        Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.1.1 2. Average Power Range Monitor - cont.
: c. Fixed Neutron    15.1.6              Inadvertent RHR            1) Automatic Initiation Flux - High                          Shutdown Cooling              - APRM Fixed Neutron Flux - High Operation                    - APRM Neutron Flux High, Setdown
                                                                              - IRM Neutron Flux - High
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown 6.3                Main Steam Line            1) Automatic Initiation 15.2.4              Isolation Valve (MSLIV) -      - Main Steam Isolation Valve - Closure Closure
* Reactor Steam Dome Pressure
                                                                              - APRM Fixed Neutron Flux - High
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                              - Reactor Pressure High ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown 15.2.11            Loss of Stator Cooling    1) Automatic Initiation
                                                                              - Reactor Vessel Steam Dome Pressure High Trip
                                                                              - APRM Fixed Neutron Flux - High
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown 15.4.5              Recirculation Flow Control 1) Automatic Initiation Failure with Increasing      - APRM Fixed Neutron Flux -High Flow                        - Reactor Water Level - High (level 8) Feedwater
                                                                            - Reactor Water Level - High (level 8) RPS
* Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown Page 5
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity RPS Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.1.1 2. Average Power Range Monitor - cont.
: c. Fixed Neutron      15.4.9              Control Rod Drop        1) Automatic Initiation Flux - High cont.                    Accident                  - APRM Fixed Neutron Flux - High
                                                                            - APRM Neutron Flux - High, Setdown Trip
* IRM Neutron Flux - High
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown
: d. Inoperable        None                None                    1) Manual SCRAM or Reactor Mode Switch - Shutdown
: 3. Reactor Vessel    15.2.11            Loss of Stator Cooling  1) Automatic Initiation Steam Dome Pressure                                                - Reactor Vessel Steam Dome Pressure High
        - High                                                              - APRM Fixed Neutron Flux - High
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown
: 4. Reactor Vessel    15.2.7              Loss of Feedwater Flow  1) Automatic Initiation Level - Low                                                        - Reactor Vessel Water Level Low Level (Level 3)
(Level 3)                                                          - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown Page 6
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity RPS Instrumentation Redundancy/Diversity Instrument          Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function      USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.1.1 4. Reactor Vessel 6.3                Feedwater Line Break    1) Automatic Initiation Level - Low      15.2.8            Outside Containment      - Reactor Vessel Level - Low, Level 3 (Level 3) - cont. 15.6.6                                      - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                      - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                      - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                      - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                      - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                      - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                      - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                      - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                      - Reactor Vessel High - Relief / LLS
                                                                      - ADS timers
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown Page 7
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity RPS Instrumentation Redundancy/Diversity Instrument        Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function    USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.1.1 4. Reactor Vessel 6.2                Loss of Coolant        1) Automatic Initiation Level - Low      6.3                Accidents                - Reactor Vessel Level - Low, Level 3 (Level 3) - cont. 15.6.5                                      - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                      - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                      - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                      - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                      - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                      - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                      - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                      - Drywell Pressure - High (LPCI A and LPCS)
                                                                      - Drywell Pressure - High (LPCI B and LPCI C)
                                                                      - Drywell Pressure - High (HPCS)
                                                                      - Drywell Pressure - High (ADS A and E)
                                                                      - Drywell Pressure - High (ADS B and F)
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown Page 8
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity RPS Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function          USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.1.1 5. Reactor Vessel      15.1.2            Feedwater Controller    1) Automatic Initiation Level - High (Level 8)                    Failure - Maximum        - Reactor Vessel Level - High (Level 8) Feedwater Demand                    - Reactor Water Level - High (level 8) RPS
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                            - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown 15.1.3            Pressure Regulator      1) Automatic Initiation Failure - Open            - Reactor Vessel Level - High (Level 8) Feedwater
                                                                            - Reactor Water Level - High (level 8) RPS
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                            - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Main Steam Line Pressure Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown Page 9
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity RPS Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.1.1 6. Main Steam        6.3                Main Steam Line          1) Automatic Initiation Isolation Valve -    15.2.4            Isolation Valve - Closure  - Main Steam Isolation Valve - Closure Closure
* Reactor Steam Dome Pressure
                                                                            - APRM Fixed Neutron Flux - High
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                            - Reactor Pressure High ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown 15.6.4            Steam System Piping      1) Automatic Initiation Break Outside              - Main Steam Isolation Valve - Closure Containment              2) Manual SCRAM or Reactor Mode Switch - Shutdown
: 7. Drywell            None              None                      1) Automatic Initiation Pressure - High                                                      - Drywell Pressure - High
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown
: 8. SCRAM Discharge Volume Water Level - High
: a. Transmitter / Trip None              None                      1) Automatic Initiation Unit                                                              - SCRAM Discharge Volume Water Level - High
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown
: b. Float Switch      None              None                      1) Automatic Initiation
                                                                            - SCRAM Discharge Volume Water Level - High
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown Page 10
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity RPS Instrumentation Redundancy/Diversity Instrument        Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function    USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.1.1 9. Turbine Stop  15.1.2            Feedwater Controller    1) Automatic Initiation Valve Closure                    Failure - Maximum        - Reactor Vessel Level - High (Level 8) Feedwater Demand                    - Reactor Water Level - High (level 8) RPS
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                      - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                      - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                      - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                      - Reactor Pressure High - Relief/LLS
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown 15.1.3            Pressure Regulator      1) Automatic Initiation Failure - Open            - Reactor Vessel Level - High (Level 8) Feedwater
                                                                      - Reactor Water Level - High (level 8) RPS
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                      - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                      - Main Steam Line Pressure Low
                                                                      - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                      - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                      - Reactor Pressure High - Relief/LLS
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown Page 11
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity RPS Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.1.1 9. Turbine Stop Valve 15.2.3            Turbine Trip            1) Automatic Initiation Closure - cont.                                                    - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                          - Reactor Pressure High - Relief/LLS
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown 15.2.5            Loss of Condenser      1) Automatic Initiation Vacuum                    - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                          - Reactor Pressure High - Relief/LLS
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown 15.3.1            Recirculation Pump Trip 1) Automatic Initiation
                                                                          - Reactor Vessel Level - High, (Level 8) Feedwater
* Reactor Water Level - High (level 8) RPS
                                                                          - Turbine Stop Valve Closure
* Reactor Water Level - High (level 8) RPS
* Turbine Control Valve Fast Closure, Trip Oil Pressure -Low
                                                                          - Main Steam Line Pressure Low
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Main Steam Isolation Valve - Closure
                                                                          - Reactor Pressure High - Relief/LLS
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown Page 12
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity RPS Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.1.1 9. Turbine Stop Valve 15.3.3            Recirculation Pump      1) Automatic Initiation Closure - cont.                      Seizure                    - Reactor Water Level - High (level 8) Feedwater
                                                                            - Reactor Water Level - High (level 8) RPS
                                                                            - Turbine Stop Valve Closure
                                                                            - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Main Steam Line Pressure Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                            - Reactor Vessel High - Relief / LLS
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown 15.3.4            Recirculation Pump Shaft 1) Automatic Initiation Break                      - Reactor Water Level - High (level 8) Feedwater
* Reactor Water Level - High (level 8) RPS
                                                                            - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Reactor Vessel High - Relief/LLS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                            - Main Steam Line Pressure Low
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown Page 13
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity RPS Instrumentation Redundancy/Diversity Instrument          Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function      USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.1.1 10. Turbine Control  15.2.2            Generator Load Rejection 1) Automatic Initiation Valve Fast                                                      - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Closure, Trip Oil
* Turbine Stop Valve Closure Pressure - Low
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                          - Reactor Pressure High - Relief/LLS
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown 15.2.6            Loss of AC power        1) Automatic Initiation
                                                                          - Reactor Vessel Level - High, (Level 8) Feedwater
* Reactor Water Level - High (level 8) RPS
* Turbine Stop Valve Closure
                                                                          - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                          - Main Steam Isolation Valve - Closure
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                          - Condenser Vacuum - Low
                                                                          - Reactor Pressure High - Relief/LLS
: 2) Manual SCRAM or Reactor Mode Switch - Shutdown
: 11. Reactor Mode    None              None Switch - Shutdown                                            1) Manual SCRAM Position
: 12. Manual SCRAM    None              None                    ---
Page 14
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
: 2.      End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation
 
==Reference:==
TS LCO 3.3.4.1 and TS B.3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT)
Instrumentation 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 MCPR Safety Limits (SLs).
The need for the additional negative reactivity in excess of that normally inserted on a scram reflects end of cycle reactivity considerations.
Flux shapes at the end of cycle are such that the control rods may not be able to ensure that thermal limits are maintained by inserting sufficient negative reactivity during the first few feet of rod travel upon a scram caused by Turbine Control Valve (TCV) Fast Closure, Trip Oil Pressure-Low, or Turbine Stop Valve (TSV) Closure.
The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity at a faster rate than the control rods can add negative reactivity.
The EOC-RPT system is a two-out-of-four logic for each Function; thus, either two TSV Closure or two TCV Fast Closure, Trip Oil Pressure-Low signals are required to actuate tripping both recirculation pumps from fast speed operation. There are two EOC-RPT breakers in series per recirculation pump. A trip in Division 1 (or 4) will cause a trip of the 'A' recirculation pump. A trip in Division 2 (or 3) will cause a trip of the 'B' recirculation pump. Both EOC-RPT breakers for each recirculation pump trip upon actuation of the EOC-RPT system. Placing an EOC-RPT bypass switch in "bypass" will allow the EOC-RPT trip capability to be maintained, When the EOC-RPT breakers trip open, the safety function is completed. The recirculation pumps may coast to stop or downshift to slow speed. Negative reactivity is provided in either case.
The TSV Closure and the TCV Fast Closure, Trip Oil Pressure-Low Functions are designed to trip the recirculation pumps from fast speed operation in the event of a turbine trip or generator load rejection to mitigate the neutron flux, heat flux, and pressure transients, and to increase the margin to the MCPR SL.
Page 15
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function          USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.4.1 a. Turbine Stop Valve  15.1.2            Feedwater Controller    1) Automatic Initiation (TSV) Closure                          Failure - Maximum          - Reactor Vessel Level - High (Level 8) Feedwater Demand                    - Reactor Water Level - High (level 8) RPS
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                              - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual RR pump Trip 15.2.3            Turbine Trip            1) Automatic Initiation
                                                                              - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual RR Pump Trip Page 16
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function          USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.4.1 a. Turbine Stop Valve  15.2.5            Loss of Condenser        1) Automatic Initiation (TSV) Closure -                        Vacuum                    - Turbine Stop Valve Closure cont.
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual RR pump trip
: b. Turbine Control      15.2.2            Generator Load Rejection 1) Automatic Initiation Valve (TCV) Fast                                                  - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Closure, Trip Oil
* Turbine Stop Valve Closure Pressure-Low
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual RR pump trip Page 17
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function          USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.4.1 b. Turbine Control      15.2.6            Loss of AC power        1) Automatic Initiation Valve (TCV) Fast                                                  - Reactor Vessel Level - High, (Level 8) Feedwater Closure, Trip Oil
* Reactor Water Level - High (level 8) RPS Pressure-Low cont.
* Turbine Stop Valve Closure
                                                                              - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                              - Main Steam Isolation Valve - Closure
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Condenser Vacuum - Low
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual RR pump trip Page 18
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
: 3.      Emergency Core Cooling System (ECCS) Instrumentation
 
==Reference:==
TS LCO 3.3.5.1 and TS B.3.3.5.1 Emergency Core Cooling System (ECCS)
Instrumentation 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).
Low Pressure Core Spray System (LPCS) may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low Low, Level 1 or Drywell Pressure-High. Each of these diverse variables is monitored by two redundant transmitters, which are, in turn, connected to two analog trip modules (ATMs). The outputs of the four ATMs (two ATMs from each of the two variables) are connected to solid state logic which is arranged in a one-out-of-two taken twice configuration. The logic can also be initiated by use of a manual push button.
Low Pressure Coolant Injection Subsystems (LPCI) is an operating mode of the Residual Heat Removal (RHR) System, with three LPCI subsystems. The LPCI subsystems may be initiated by automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low Low, Level 1 or Drywell Pressure-High. Each of these diverse variables is monitored by two redundant transmitters per Division, which are, in turn, connected to two ATMs. The outputs of the Division 2 LPCI (loops B and C) ATMs (two ATMs from each of the two variables) are connected to solid state logic which is arranged in a one-out-of-two taken twice configuration. The Division 1 LPCI (loop A) receives its initiation signal from the LPCS logic, which uses a similar one-out-of-two taken twice logic. The two Divisions can also be initiated by use of a manual push button (one per Division).
High Pressure Core Spray System (HPCS) may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low, Level 2 or Drywell Pressure-High from Division 3 or Division 4 instrumentation. The outputs of the ATMs are connected to solid state logic which is arranged in a one-out-of-two taken twice logic for each variable. The HPCS system also monitors the water levels in the Reactor Core Isolation Cooling (RCIC) Storage Tank and the suppression pool, since these are the two sources of water for HPCS operation. Reactor grade water in the RCIC Storage Tank is the normal and preferred source. Upon receipt of a HPCS initiation signal, the RCIC Storage Tank suction valve is automatically signaled to open (it is normally in the open position), unless the suppression pool suction valve is open. If the water level in the RCIC Storage Tank falls below a preselected level, first the suppression pool suction valve automatically opens, and then the RCIC Storage Tank suction valve automatically closes. The HPCS system provides makeup water to the reactor until the reactor vessel water level reaches the high water level (Level 8) trip, at which time the HPCS injection valve closes. A subsequent Reactor Vessel Water Level-Low Low, Level 2 will reopen the HPCS injection valve to provide makeup water to the reactor.
Page 19
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Automatic Depressurization System (ADS) may be initiated by either automatic or manual means. Automatic initiation occurs when signals indicating Reactor Vessel Water Level-Low Low Low, Level 1; Drywell Pressure-High or ADS Drywell Pressure Bypass Timer; confirmed Reactor Vessel Water Level-Low, Level 3; and either LPCS or LPCI Pump Discharge Pressure-High are all present, and the ADS Initiation Timer has timed out. There are two transmitters each for Reactor Vessel Water Level-Low Low Low, Level 1 and Drywell Pressure-High, and one transmitter for confirmed Reactor Vessel Water Level-Low, Level 3 in each of the two ADS trip systems. Each of these transmitters connects to an ATM, which then inputs to the solid state initiation logic.
Each ADS trip system (trip system 1 and trip system 2) includes a time delay between satisfying the initiation logic and the actuation of the ADS valves. The time delay chosen is long enough that the HPCS has time to operate to recover to a level above Level 1, yet not so long that the LPCI and LPCS systems are unable to adequately cool the fuel if the HPCS fails to maintain level. Either ADS trip system 1 or trip system 2 will cause all the ADS relief valves to open.
Once the Drywell Pressure-High or ADS initiation signals are present, they are individually sealed in until manually reset Diesel Generators (DG) for Division 1, 2, and 3 DGs may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low Low, Level 1 or Drywell Pressure-High for DGs 1A and 1B, and Reactor Vessel Water Level-Low Low, Level 2 or Drywell Pressure-High for DG 1C. The DGs are also initiated upon loss of voltage signals referred to in LCO 3.3.8.1, "Loss of Power (LOP) Instrumentation,"
The actions of the ECCS are explicitly assumed in the safety analyses of USAR section 5.2.2, Overpressure Protection; USAR section 6.3, Emergency Core Cooling System and USAR chapter 15, Accident Analysis. The ECCS is initiated to preserve the integrity of the fuel cladding by limiting the post LOCA peak cladding temperature to less than the 10 CFR 50.46 limits.
In the event of a break in a pipe that is not a part of the ECCS, no single active component failure in the ECCS shall prevent automatic initiation and successful operation of less than the following combination of ECCS equipment:
: 1. three LPCI loops, the LPCS and the ADS (i.e., HPCS failure); or
: 2. two LPCI loops, the HPCS and the ADS (i.e., "LPCS diesel generator" (i.e., Div 1 DG) failure); or
: 3. one LPCI loop, the LPCS, the HPCS and ADS (i.e., "LPCI diesel generator" (i.e., Div 2 DG) failure).
In the event of a break in a pipe that is a part of the ECCS, no single active component failure in the ECCS shall prevent automatic initiation and successful operation of less than the following combination of ECCS equipment:
: 1. two LPCI loops and the ADS; or
: 2. one LPCI loop, the LPCS and the ADS; or
: 3. one LPCI loop, the HPCS and the ADS; or
: 4. the LPCS, the HPCS and ADS.
Page 20
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 1. Low Pressure Coolant Injection-A (LPCI) and Low Pressure Core Spray (LPCS) Subsystems
: a. Reactor Vessel    6.3                Feedwater Line Break    1) Automatic Initiation Water Level-Low  15.2.8              Outside Containment        - Reactor Vessel Level - Low, Level 3 Low Low, Level 1  15.6.6                                        - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                            - Reactor Vessel High - Relief / LLS
                                                                            - ADS timers
: 2) Manual LPCI A or LPCS initiation 6.3                Steam System Piping      1) Automatic Initiation 15.6.4            Break Outside              - Main Steam Line Flow - High Containment                - Main Steam Isolation Valve - Closure
                                                                            - Reactor Pressure High - Relief/LLS
                                                                            - Reactor Vessel Water Level - Low Low Low, Level 1
                                                                            - ADS Initiation Timer
: 2) Manual LPCI A or LPCS initiation Page 21
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 1. Low Pressure Coolant Injection-A (LPCI) and Low Pressure Core Spray (LPCS) Subsystems - cont.
: a. Reactor Vessel    6.2                Loss of Coolant Accident 1) Automatic Initiation Water Level-Low    6.3                                            - Reactor Vessel Level - Low, Level 3 Low Low, Level 1  15.6.5
* Drywell pressure - High (RPS) cont.                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                            - Drywell Pressure - High (LPCI A and LPCS)
                                                                            - Drywell Pressure - High (LPCI B and LPCI C)
                                                                            - Drywell Pressure - High (HPCS)
                                                                            - Drywell Pressure - High (ADS A and E)
                                                                            - Drywell Pressure - High (ADS B and F)
: 2) Manual LPCI A or LPCS initiation Page 22
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function      USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 1. Low Pressure Coolant Injection-A (LPCI) and Low Pressure Core Spray (LPCS) Subsystems - cont.
: b. Drywell (DW)      6.2                Loss of Coolant Accident 1) Automatic Initiation Pressure-High      6.3                                            - Reactor Vessel Level - Low, Level 3 15.6.5
* Drywell pressure - High (RPS)
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                            - Drywell Pressure - High (LPCI A and LPCS)
                                                                            - Drywell Pressure - High (LPCI B and LPCI C)
                                                                            - Drywell Pressure - High (HPCS)
                                                                            - Drywell Pressure - High (ADS A and E)
                                                                            - Drywell Pressure - High (ADS B and F)
: 2) Manual LPCI A or LPCS initiation
: c. LPCI Pump A        None              None                    1) LPCS initiation Start-Time Delay Logic Card Implicit to LPCI A function Page 23
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function      USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1  1. Low Pressure Coolant Injection-A (LPCI) and Low Pressure Core Spray (LPCS) Subsystems - cont.
: d. Reactor Vessel      None              None                      1) One out of two logic Pressure-Low (Injection Permissive)
Implicit to LPCS or LPCI A function
: e. LPCS Pump          None              None                      1) LPCI A Discharge Flow-                                                2) LPCI B Low (Bypass)                                                    3) LPCI C Implicit to LPCS function
: f. LPCI Pump A        None              None                      1) LPCS Discharge Flow-                                                2) LPCI B Low (Bypass)                                                    3) LPCI C Implicit to LPCI A function
: g. Manual Initiation  None              None                        ---
Page 24
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 2. LPCI B and LPCI C Subsystems
: a. Reactor Vessel    6.3                Feedwater Piping Break 1) Automatic Initiation Water Level-Low    15.2.8            Feedwater Line Break      - Reactor Vessel Level - Low, Level 3 Low Low, Level 1  15.6.6            Outside Containment      - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                          - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                          - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                          - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                          - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                          - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                          - Reactor Vessel High - Relief / LLS
                                                                          - ADS timers
: 2) Manual LPCI B and LPCI B initiation 6.3                Steam System Piping    1) Automatic Initiation 15.6.4            Break Outside            - Main Steam Line Flow - High Containment              - Main Steam Isolation Valve - Closure
                                                                          - Reactor Pressure High - Relief/LLS
                                                                          - Reactor Vessel Water Level - Low Low Low, Level 1
                                                                          - ADS Initiation Timer
: 2) Manual LPCI B and LPCI B initiation Page 25
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 2. LPCI B and LPCI C Subsystems - cont.
: a. Reactor Vessel    6.2                Loss of Coolant Accident 1) Automatic Initiation Water Level-Low    6.3                                            - Reactor Vessel Level - Low, Level 3 Low Low, Level 1 - 15.6.5
* Drywell pressure - High (RPS) cont.                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                            - Drywell Pressure - High (LPCI A and LPCS)
                                                                            - Drywell Pressure - High (LPCI B and LPCI C)
                                                                            - Drywell Pressure - High (HPCS)
                                                                            - Drywell Pressure - High (ADS A and E)
                                                                            - Drywell Pressure - High (ADS B and F)
: 2) Manual LPCI B and LPCI B initiation
: b. Drywell Pressure-  6.2              Loss of Coolant Accident  1) Same as above High              6.3 15.6.5 Page 26
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function      USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1  2. LPCI B and LPCI C Subsystems - cont.
: c. LPCI Pump B        None              None                    1) LPCI A system Start-Time Delay                                              2) LPCI C system Logic Card                                                    3) LPCS system Implicit to LPCI B function
: d. Reactor Vessel      None              None                    1) RHR B Reactor pressure Pressure-Low                                                  2) RPS Reactor pressure (Injection Permissive)
Implicit to LPCI B and LPCI C function
: e. LPCI Pump B LPCI None                None                    1) LPCI A Pump C Discharge                                              2) LPCS Flow-Low (Bypass)
Implicit to LPCI B and LPCI C function
: f. Manual Initiation  None              None                      ---
Page 27
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 3. High Pressure Core Spray (HPCS) System
: a. Reactor Vessel    15.1.1            Loss of Feedwater      1) Automatic Initiation Water Level-                          Heating                  - APRM Flow Biased Simulated Thermal Power - Upscale Low Low,
* APRM Fixed Neutron Flux - High Level 2                                                        - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
: 2) Manual HPCS initiation 15.1.2              Feedwater Controller  1) Automatic Initiation Failure - Maximum        - Reactor Vessel Level - High (Level 8) Feedwater Demand                    - Reactor Water Level - High (level 8) RPS
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                          - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                          - Reactor Pressure High - Relief/LLS
: 2) Manual HPCS initiation Page 28
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 3. High Pressure Core Spray (HPCS) System - cont.
: a. Reactor Vessel    15.1.3              Pressure Regulator        1) Automatic Initiation Water Level-                          Failure - Open              - Reactor Vessel Level - High (Level 8) Feedwater Low Low,                                                          - Reactor Water Level - High (level 8) RPS Level 2 - cont.
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                              - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                              - Main Steam Line Pressure Low
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual HPCS initiation 6.3                Main Steam Line          1) Automatic Initiation 15.2.4            Isolation Valve (MSLIV) -    - Main Steam Isolation Valve - Closure Closure
* Reactor Steam Dome Pressure
                                                                              - APRM Fixed Neutron Flux - High
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                              - Reactor Pressure High ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual HPCS initiation Page 29
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 3. High Pressure Core Spray (HPCS) System - cont.
: a. Reactor Vessel    6.3                Loss of Condenser      1) Automatic Initiation Water Level-Low    15.2.5            Vacuum                    - Turbine Stop Valve Closure Low, Level 2 -
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low cont.
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                          - Reactor Pressure High - Relief/LLS
: 2) Manual HPCS initiation 6.3                Loss of AC Power      1) Automatic Initiation 15.2.6                                      - Reactor Vessel Level - High, (Level 8) Feedwater
* Reactor Water Level - High (level 8) RPS
* Turbine Stop Valve Closure
                                                                          - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                          - Main Steam Isolation Valve - Closure
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                          - Condenser Vacuum - Low
                                                                          - Reactor Pressure High - Relief/LLS
: 2) Manual HPCS initiation Page 30
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 3. High Pressure Core Spray (HPCS) System - cont.
: a. Reactor Vessel    6.3                Loss of Feedwater Flow 1) Automatic Initiation Water Level-Low 15.2.7                                        - Reactor Vessel Water Level Low Level (Level 3)
Low, Level 2 -                                                - Reactor Vessel Water Level Low Low Level (Level 2) RCIC cont.                                                          - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
: 2) Manual HPCS initiation 6.3                Feedwater Piping Break 1) Automatic Initiation 15.2.8            Feedwater Line Break      - Reactor Vessel Level - Low, Level 3 15.6.6            Outside Containment      - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                          - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                          - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                          - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                          - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                          - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                          - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                          - Reactor Vessel High - Relief / LLS
                                                                          - ADS timers
: 2) Manual HPCS initiation Page 31
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 3. High Pressure Core Spray (HPCS) System - cont.
: a. Reactor Vessel    6.3                Recirculation Pump Trip 1) Automatic Initiation water level Water Level-Low    15.3.1                                        - Reactor Vessel Level - High, (Level 8) Feedwater Low, Level 2 -
* Reactor Water Level - High (level 8) RPS cont.                                                            - Turbine Stop Valve Closure
* Reactor Water Level - High (level 8) RPS
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Main Steam Line Pressure Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Main Steam Isolation Valve - Closure
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual HPCS initiation 6.3                Recirculation Pump      1) Automatic Initiation water level 15.3.3            Seizure                    - Reactor Water Level - High (level 8) Feedwater
                                                                            - Reactor Water Level - High (level 8) RPS
                                                                            - Turbine Stop Valve Closure
                                                                            - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Main Steam Line Pressure Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                            - Reactor Vessel High - Relief / LLS
: 2) Manual HPCS initiation Page 32
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 3. High Pressure Core Spray (HPCS) System - cont.
: a. Reactor Vessel    6.3                Recirculation Pump Shaft 1) Automatic Initiation water level Water Level-Low    15.3.4            Break                      - Reactor Water Level - High (level 8) Feedwater Low, Level 2 -
* Reactor Water Level - High (level 8) RPS cont.                                                            - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Reactor Vessel High - Relief / LLS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                            - Main Steam Line Pressure Low
: 2) Manual HPCS initiation 6.3                Loss of Coolant Accident 1) Automatic Initiation water level or DW press 15.6.5                                        - Reactor Vessel Level - Low, Level 3
* Drywell pressure - High (RPS)
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                            - Drywell Pressure - High (LPCI A and LPCS)
                                                                            - Drywell Pressure - High (LPCI B and LPCI C)
                                                                            - Drywell Pressure - High (HPCS)
                                                                            - Drywell Pressure - High (ADS A and E)
                                                                            - Drywell Pressure - High (ADS B and F)
: 2) Manual HPCS initiation Page 33
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 3. High Pressure Core Spray (HPCS) System - cont.
: a. Reactor Vessel    6.3                Feedwater Line Break-    1) Automatic Initiation water level or DW press Water Level-Low      15.2.8            Outside Containment        - Reactor Vessel Level - Low, Level 3 Low, Level 2 - cont  15.6.6                                        - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                            - Reactor Vessel High - Relief / LLS
                                                                            - ADS timers
: 2) Manual HPCS initiation
: b. Drywell Pressure- 6.3                Loss of Coolant Accident 1) Automatic Initiation water level or DW press High              15.6.5                                        - Reactor Vessel Level - Low, Level 3
* Drywell pressure - High (RPS)
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                            - Drywell Pressure - High (LPCI A and LPCS)
                                                                            - Drywell Pressure - High (LPCI B and LPCI C)
                                                                            - Drywell Pressure - High (HPCS)
                                                                            - Drywell Pressure - High (ADS A and E)
                                                                            - Drywell Pressure - High (ADS B and F)
: 2) Manual HPCS initiation Page 34
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function      USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 3. High Pressure Core Spray (HPCS) System - cont.
: c. Reactor Vessel      None              None                    1) Two out of Two logic Water Level-High, Level 8 Implicit to HPCS function
: d. RCIC Storage Tank None                None                    1) One out of Two logic Level-Low Implicit to HPCS function
: e. Suppression Pool    None              None                    1) One out of Two logic Water Level-High Implicit to HPCS function
: f. HPCS Pump          None              None                    1) RCIC Discharge Pressure-High (Bypass)
Implicit to HPCS function
: g. HPCS System        None              None                    1) RCIC Flow Rate-Low (Bypass)
Implicit to HPCS function
: h. Manual Initiation  None              None                      ---
Page 35
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 4. Automatic Depressurization System (ADS) Trip System 1 (Logic A and E)
: a. Reactor Vessel    6.3                Feedwater Piping Break  1) Automatic Initiation Water Level-      15.2.8            Feedwater Line Break        - Reactor Vessel Level - Low, Level 3 Low Low Low,      15.6.6            Outside Containment          - Reactor Vessel Water Level Low Low Level (Level 2) RCIC Level 1                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Reactor Vessel High - Relief / LLS
                                                                              - ADS timers
: 2) Manual ADS initiation 6.3                Steam System Piping      1) Automatic Initiation 15.6.4            Break Outside                - Main Steam Line Flow - High Containment                  - Main Steam Isolation Valve - Closure
                                                                              - Reactor Pressure High - Relief/LLS
                                                                              - Reactor Vessel Water Level - Low Low Low, Level 1
                                                                              - ADS Initiation Timer
: 2) Manual ADS initiation Page 36
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 4. Automatic Depressurization System (ADS) Trip System 1 (Logic A and E)
: a. Reactor Vessel    6.2                Loss of Coolant Accident 1) Automatic Initiation Water Level-      6.3                                            - Reactor Vessel Level - Low, Level 3 Low Low Low,      15.6.5
* Drywell pressure - High (RPS)
Level 1 - cont.                                                    - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Drywell Pressure - High (LPCI A and LPCS)
                                                                              - Drywell Pressure - High (LPCI B and LPCI C)
                                                                              - Drywell Pressure - High (HPCS)
                                                                              - Drywell Pressure - High (ADS A and E)
                                                                              - Drywell Pressure - High (ADS B and F)
: 2) Manual ADS initiation Page 37
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 4. Automatic Depressurization System (ADS) Trip System 1 (Logic A and E) - cont.
: b. Drywell Pressure-  6.3                Loss of Coolant Accident 1) Automatic Initiation High              15.6.5                                          - Reactor Vessel Level - Low, Level 3
* Drywell pressure - High (RPS)
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Drywell Pressure - High (LPCI A and LPCS)
                                                                              - Drywell Pressure - High (LPCI B and LPCI C)
                                                                              - Drywell Pressure - High (HPCS)
                                                                              - Drywell Pressure - High (ADS A and E)
                                                                              - Drywell Pressure - High (ADS B and F)
: 2) Manual ADS initiation Page 38
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 4. Automatic Depressurization System (ADS) Trip System 1 (Logic A and E) - cont.
: c. ADS Initiation      6.3                Feedwater Piping Break 1) Automatic Initiation Timer                15.2.8            Feedwater Line Break      - Reactor Vessel Level - Low, Level 3 15.6.6                                      - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Reactor Vessel High - Relief / LLS
                                                                              - ADS timers
: 2) Manual ADS initiation 6.3                Steam System Piping    1) Automatic Initiation 15.6.4            Break                    - Main Steam Line Flow - High
                                                                              - Main Steam Isolation Valve - Closure
                                                                              - Reactor Pressure High - Relief/LLS
                                                                              - Reactor Vessel Water Level - Low Low Low, Level 1
                                                                              - ADS Initiation Timer
: 2) Manual ADS initiation
: d. Reactor Vessel        None              None                  1) Automatic Initiation Water Level-Low,                                                  - ADS Trip System (Logic B and F)
Level 3                                                        2) Manual ADS initiation (Confirmatory)
Implicit to ADS function
: e. LPCS Pump            None              None                  1) Automatic Initiation Discharge                                                        - LPCI A Pressure-High Implicit to ADS function Page 39
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 4. Automatic Depressurization System (ADS) Trip System 1 (Logic A and E) - cont.
: f. LPCI Pump A          None              None                  1) Automatic Initiation Discharge                                                          - LPCS Pressure- High Implicit to ADS function
: g. ADS Drywell          6.3                Feedwater Line Break  1) Automatic Initiation Pressure Bypass      15.2.8            Outside Containment        - Reactor Vessel Level - Low, Level 3 Timer                15.6.6                                        - Reactor Vessel Water Level Low Low Level (Level 2) RCIC Implicit to ADS                                                      - Reactor Vessel Water Level Low Low Level (Level 2) HPCS function. If the event                                                - Reactor Vessel Water Level Low Low Level (Level 2) ATWS occurs outside of the                                                    RPT Drywell, Hi Drywell
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS pressure may not be                                                      Alternate Rod Insertion (ARI) present and is                                                        - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E) bypassed to allow for                                                - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
ADS initiation.                                                      - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Reactor Vessel High - Relief / LLS
                                                                              - ADS timers
: 2) Manual SCRAM/ECCS/ISOL
: h. Manual Initiation    None              None                    ---
Page 40
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 5. Automatic Depressurization System (ADS) Trip System 2 (Logic B and F)
: a. Reactor Vessel    6.3                Feedwater Piping Break  1) Automatic Initiation Water Level-Low    15.2.8            Feedwater Line Break        - Reactor Vessel Level - Low, Level 3 Low Low, Level 1  15.6.6            Outside Containment          - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Reactor Vessel High - Relief / LLS
                                                                              - ADS timers
: 2) Manual ADS initiation 6.3                Steam System Piping      1) Automatic Initiation 15.6.4            Break Outside                - Main Steam Line Flow - High Containment                  - Main Steam Isolation Valve - Closure
                                                                              - Reactor Pressure High - Relief/LLS
                                                                              - Reactor Vessel Water Level - Low Low Low, Level 1
                                                                              - ADS Initiation Timer
: 2) Manual ADS initiation Page 41
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 5. Automatic Depressurization System (ADS) Trip System 2 (Logic B and F)
: a. Reactor Vessel    6.3                Loss of Coolant Accident 1) Automatic Initiation Water Level-Low    15.6.5                                          - Reactor Vessel Level - Low, Level 3 Low Low, Level 1 -
* Drywell pressure - High (RPS) cont.                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Drywell Pressure - High (LPCI A and LPCS)
                                                                              - Drywell Pressure - High (LPCI B and LPCI C)
                                                                              - Drywell Pressure - High (HPCS)
                                                                              - Drywell Pressure - High (ADS A and E)
                                                                              - Drywell Pressure - High (ADS B and F)
: 2) Manual ADS initiation Page 42
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 5. Automatic Depressurization System (ADS) Trip System 2 (Logic B and F) - cont.
: b. Drywell Pressure-  6.3                Loss of Coolant Accident 1) Automatic Initiation High              15.6.5                                          - Reactor Vessel Level - Low, Level 3
* Drywell pressure - High (RPS)
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Drywell Pressure - High (LPCI A and LPCS)
                                                                              - Drywell Pressure - High (LPCI B and LPCI C)
                                                                              - Drywell Pressure - High (HPCS)
                                                                              - Drywell Pressure - High (ADS A and E)
                                                                              - Drywell Pressure - High (ADS B and F)
: 2) Manual ADS initiation Page 43
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1  5. Automatic Depressurization System (ADS) Trip System 2 (Logic B and F) - cont.
: c. ADS Initiation      6.3                Feedwater Piping Break 1) Automatic Initiation Timer                15.2.8            Feedwater Line Break      - Reactor Vessel Level - Low, Level 3 15.6.6                                        - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Reactor Vessel High - Relief / LLS
                                                                              - ADS timers
: 2) Manual ADS initiation 6.3                Steam System Piping    1) Automatic Initiation 15.6.4            Break                      - Main Steam Line Flow - High
                                                                              - Main Steam Isolation Valve - Closure
                                                                              - Reactor Pressure High - Relief/LLS
                                                                              - Reactor Vessel Water Level - Low Low Low, Level 1
                                                                              - ADS Initiation Timer
: 2) Manual ADS initiation
: d. Reactor Vessel      None              None                  1) Automatic Initiation Water Level-Low,                                                  - ADS Trip System (Logic A and E)
Level 3                                                        2) Manual ADS initiation (Confirmatory)
Implicit to ADS function
: e. LPCS Pump            None              None                  1) Automatic Initiation Discharge                                                          - LPCI B and C Pressure-High                                                  2) Manual ADS initiation Implicit to ADS function Page 44
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Emergency Core Cooling System (ECCS) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.1 5. Automatic Depressurization System (ADS) Trip System 2 (Logic B and F) - cont.
: f. LPCI Pump B & C      None              None                  1) Automatic Initiation Discharge                                                          - LPCS Pressure-High                                                  2) Manual ADS initiation Implicit to ADS function
: g. ADS Drywell          6.3                Feedwater Line Break  1) Automatic Initiation Pressure Bypass      15.2.8            Outside Containment        - Reactor Vessel Level - Low, Level 3 Timer                15.6.6                                        - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS Implicit to ADS                                                      - Reactor Vessel Water Level Low Low Level (Level 2) ATWS function. If the event                                                  RPT occurs outside of the
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Drywell, Hi Drywell                                                      Alternate Rod Insertion (ARI) pressure may not be                                                  - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E) present and is                                                        - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F) bypassed to allow for                                                - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and ADS initiation.                                                          LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Reactor Vessel High - Relief / LLS
                                                                              - ADS timers
: 2) Manual SCRAM/ECCS/ISOL
: h. Manual Initiation    None              None                  ---
Page 45
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
: 4.      Reactor Core Isolation Cooling (RCIC) System
 
==Reference:==
TS LCO 3.3.5.3 and TS B3.3.5.3 Reactor Core Isolation Cooling (RCIC) System Instrumentation The RCIC System instrumentation is to initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink (the main condenser) and normal coolant makeup flow from the Reactor Feedwater System is unavailable, such that initiation of the low pressure Emergency Core Cooling Systems (ECCS) pumps does not occur.
The RCIC System may be initiated by either automatic or manual means. Automatic initiation occurs for conditions of Reactor Vessel Water Level-Low Low, Level 2 arranged in a one-out-of-two taken twice logic configuration.
The RCIC System also monitors the water levels in the RCIC Storage Tank (2 transmitters) and the suppression pool (2 transmitters), since these are the two sources of water for RCIC operation. Reactor grade water in the RCIC Storage Tank is the normal source. Any one of the transmitters can cause an automatic shift to the suppression pool if RCIC tank level is low or if suppression pool level is high.
The RCIC System provides makeup water to the reactor until the reactor vessel water level reaches the high water level (Level 8) trip (two-out-of-two logic), at which time the RCIC steam supply, and cooling water supply valves close (the injection valve also closes due to the closure of the steam supply valve). The RCIC System restarts if vessel level again drops to the low level initiation point (Level 2).
Page 46
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Reactor Core Isolation Cooling (RCIC) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function      USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.3 1. Reactor Vessel    15.1.1            Loss of Feedwater Heating 1) Automatic Initiation Water Level Low Low                                                  - APRM Flow Biased Simulated Thermal Power - Upscale (Level 2)
* APRM Fixed Neutron Flux - High
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
: 2) Manual RCIC initiation 15.1.2            Feedwater Controller      1) Automatic Initiation water level Failure - Maximum            - Reactor Vessel Level - High (Level 8) Feedwater Demand                      - Reactor Water Level - High (level 8) RPS
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                            - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Pressure High- Relief/LLS
: 2) Manual RCIC initiation Page 47
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Reactor Core Isolation Cooling (RCIC) Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function      USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.3 1. Reactor Vessel    15.1.3            Pressure Regulator        1) Automatic Initiation Water Level Low Low                    Failure - Open              - Reactor Vessel Level - High (Level 8) Feedwater (Level 2) - cont                                                    - Reactor Water Level - High (level 8) RPS
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                            - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Main Steam Line Pressure Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual RCIC Initiation 6.3                Main Steam Line Isolation 1) Automatic Initiation 15.2.4              Valve (MSLIV) - Closure      - Main Steam Isolation Valve - Closure
* Reactor Steam Dome Pressure
                                                                            - APRM Fixed Neutron Flux - High
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                            - Reactor Pressure High ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual RCIC Initiation Page 48
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Reactor Core Isolation Cooling (RCIC) Instrumentation Redundancy/Diversity Instrument          Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function      USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.3 1. Reactor Vessel    15.2.5            Loss of Condenser        1) Automatic Initiation Water Level Low Low                    Vacuum                      - Turbine Stop Valve Closure (Level 2) - cont
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual RCIC initiation 15.2.6            Loss of AC Power        1) Automatic Initiation water level
                                                                            - Reactor Vessel Level - High, (Level 8) Feedwater
* Reactor Water Level - High (level 8) RPS
* Turbine Stop Valve Closure
                                                                            - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Main Steam Isolation Valve - Closure
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Condenser Vacuum - Low
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual RCIC initiation Page 49
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Reactor Core Isolation Cooling (RCIC) Instrumentation Redundancy/Diversity Instrument          Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function      USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.3 1. Reactor Vessel    15.2.7            Loss of Feedwater        1) Automatic Initiation Water Level Low Low                    Flow                        - Reactor Vessel Water Level Low Level (Level 3)
(Level 2) - cont                                                    - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
: 2) Manual RCIC initiation 6.3                Feedwater Line          1) Automatic Initiation 15.2.8              Break Outside              - Reactor Vessel Level - Low, Level 3 15.6.6              Containment                - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                            - Reactor Vessel High - Relief / LLS
                                                                            - ADS timers
: 2) Manual RCIC initiation Page 50
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Reactor Core Isolation Cooling (RCIC) Instrumentation Redundancy/Diversity Instrument          Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function      USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.3 1. Reactor Vessel  6.3                Feedwater Line            1) Automatic Initiation Water Level Low Low 15.2.8            Break Outside                - Reactor Vessel Level - Low, Level 3 (Level 2) - cont    15.6.6            Containment                  - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                            - Reactor Vessel High - Relief / LLS
                                                                            - ADS timers
: 2) Manual RCIC initiation 15.3.1            Recirculation            1) Automatic Initiation Pump Trip                    - Reactor Vessel Level - High, (Level 8) Feedwater
* Reactor Water Level - High (level 8) RPS
                                                                            - Turbine Stop Valve Closure
* Reactor Water Level - High (level 8) RPS
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Main Steam Line Pressure Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Main Steam Isolation Valve - Closure
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual RCIC initiation Page 51
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Reactor Core Isolation Cooling (RCIC) Instrumentation Redundancy/Diversity Instrument          Credited Safety Analysis Event                        Redundant/Diverse Instrumentation TS Function      USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.3 1. Reactor Vessel  15.3.3            Recirculation            1) Automatic Initiation Water Level Low Low                    Pump Seizure                  - Reactor Water Level - High (level 8) Feedwater (Level 2) - cont                                                    - Reactor Water Level - High (level 8) RPS
                                                                            - Turbine Stop Valve Closure
                                                                            - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Main Steam Line Pressure Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                            - Reactor Vessel High - Relief / LLS
: 2) Manual RCIC initiation 15.3.4            Recirculation              1) Automatic Initiation Pump Shaft Break              - Reactor Water Level - High (level 8) Feedwater
* Reactor Water Level - High (level 8) RPS
                                                                            - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Reactor Vessel High - Relief / LLS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                            - Main Steam Line Pressure Low
: 2) Manual RCIC initiation Page 52
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Reactor Core Isolation Cooling (RCIC) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                        Redundant/Diverse Instrumentation TS Function          USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.5.3 1. Reactor Vessel        6.2                Loss of Coolant          1) Automatic Initiation Water Level Low Low      6.3                Accidents                    - Reactor Vessel Level - Low, Level 3 (Level 2) - cont          15.6.5
* Drywell pressure - High (RPS)
                                                                                  - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                                  - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                                  - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                                  - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                                  - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                                  - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                                  - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                                  - Drywell Pressure - High (LPCI A and LPCS)
                                                                                  - Drywell Pressure - High (LPCI B and LPCI C)
                                                                                  - Drywell Pressure - High (HPCS)
                                                                                  - Drywell Pressure - High (ADS A and E)
                                                                                  - Drywell Pressure - High (ADS B and F)
: 2) Manual SCRAM/ECCS/ISOL
: 2. Reactor Vessel        None              None                      Two out of Two logic Water Level - High (Level 8)
Implicit to RCIC function
: 3. RCIC Storage Tank None                    None                      One out of Two logic
        - Low Implicit to RCIC function
: 4. Suppression Pool      None              None                      One out of Two logic Water Level-High Implicit to RCIC function
: 5. Manual Initiation      None              None                      ---
Page 53
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
: 5.      Primary Containment and Drywell Isolation Instrumentation
 
==Reference:==
TS LCO 3.3.6.1 and TS B3.3.6.1 Primary Containment and Drywell Isolation Instrumentation The primary containment and drywell 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 Design Basis Accidents (DBAs).
Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation logic are:
(a) reactor vessel water level (b) ambient temperatures (c) main steam line (MSL) flow measurement (d) Standby Liquid Control (SLC) System initiation (e) condenser vacuum loss (f) main steam line pressure (g) reactor core isolation cooling (RCIC) steam line flow (h) ventilation exhaust radiation (i) RCIC steam line pressure (j) RCIC turbine exhaust diaphragm pressure (k) reactor water cleanup (RWCU) differential flow (l) reactor steam dome pressure (m) drywell pressure and (n) containment pressure The primary containment and drywell isolation instrumentation has inputs to the trip logic from the isolation Functions listed below.
: 1. Main Steam Line Isolation
* The four channels input to four separate two-out-of-four logic divisions.
o The outputs from these logic divisions are combined into two two-out-of-two logic trip systems to isolate all main steam isolation valves (MSIVs) and MSL drain valves.
* The exception to this arrangement is the Main Steam Line Flow-High Function. This Function uses 16 flow channels, four for each steam line.
o The four flow channels associated with a steam line are combined in a two-out-of-four logic configuration.
o The outputs of the high steam flow logic for each of the steam lines are combined in the two two-out-of- two logic trip systems described above.
Page 54
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
: 2. Primary Containment and Drywell Isolation
* Each Primary Containment Isolation and Drywell Function receives inputs from four channels.
* The outputs from these channels are arranged into two logic trip systems.
o One trip system initiates isolation of all inboard PCIVs and drywell isolation valves, while the other trip system initiates isolation of all outboard PCIVs and drywell isolation valves.
o Each trip system logic closes one of the two valves on each penetration so that operation of either trip system isolates the penetration.
o This logic configuration also provides automatic actuation capability for the Division 1 and 2 Shutdown Service Water (SX) subsystems.
: 3. Reactor Core Isolation Cooling System Isolation
* Most Functions receive input from two channels, with each channel in one trip system.
o Each of the two trip systems is connected to one of the two valves on each RCIC penetration so that operation of either trip system isolates the penetration.
The exception to this arrangement is the RCIC Turbine Exhaust Diaphragm Pressure - High Function.
* The Reactor Vessel Water Level - Low Low, Level 2 RCIC initiation function receives inputs from four channels.
o The outputs from these channels are arranged into two logic trip systems.
Each trip system logic closes one of the two valves on the RCIC penetration so that operation of either trip system isolates the penetration.
* The RCIC Turbine Exhaust Diaphragm Pressure - High Function receives input from four turbine exhaust diaphragm pressure channels.
o The outputs from the turbine exhaust diaphragm pressure channels are connected into two two-out-of-two trip systems, each trip system isolating two RCIC valves.
* There is one manual isolation switch which can isolate only the outboard RCIC System containment isolation valves.
: 4. Reactor Water Cleanup System Isolation
* Most Functions receive input from two channels with each channel in one trip system using one-out-of-one logic.
* Functions 4.c and 4.d (RWCU Heat Exchanger Room Temperature and RWCU Pump Room Temperature) have one channel in each trip system in each room for a total of four channels for Function 4.c and six channels for Function 4.d, but the logic is the same (one-out-of-one).
Page 55
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
: 4. Reactor Water Cleanup System Isolation - cont.
* Each of the two trip systems is connected to one of the two valves on each RWCU penetration so that operation of either trip system isolates the penetration.
o The exception to this arrangement is the Reactor Vessel Water Level-Low Low, Level 2 Function. This Function receives input from four reactor vessel water level channels. The outputs from the reactor vessel water level channels are connected into two two-out-of-two trip systems, each trip system isolating one of the two RWCU valves.
: 5. RHR System Isolation
* The RHR System Isolation Function receives input signals from instrumentation for:
o Reactor Vessel Water Level-Low Low Low, Level 1; o Reactor Vessel Water Level - Low, Level 3; o Drywell Pressure - High; o Reactor Vessel Pressure - High; o RHR Equipment Room Ambient Temperature - High; and o Manual Initiation Functions.
* The Reactor Vessel Water Level-Low Low Low, Level 1; Reactor Vessel Water Level-Low, Level 3; Reactor Steam Dome Pressure-High; and Drywell Pressure-High Functions each have four channels.
* The outputs from the reactor vessel water level (level 1) and drywell pressure channels are connected in two one-out-of-two twice trip systems.
* The reactor vessel water level (level 3) is combined with the drywell pressure channels in two one-out-of-two twice trip systems and with the reactor vessel pressure channels in two one-out-of-two twice trip systems.
* The RHR Heat Exchanger Room Ambient Temperature Function receives input from four channels with each channel in one trip system in one room using one-out-of-one logic.
o Each of the two trip systems is connected to one of the two valves on each shutdown cooling penetration so that operation of either trip system isolates the penetration.
The isolation signals generated by the primary containment and drywell isolation instrumentation are implicitly assumed in the safety analyses of USAR Chapter 6.2 and USAR Chapter 15 to initiate closure of PCIVs to limit offsite doses.
Containment and Reactor Vessel Isolation Control System Instrumentation and Controls initiate closure of various automatic isolation valves if monitored system variables exceed pre-established limits. This action limits the loss of coolant from the reactor coolant pressure boundary and the release of radioactive materials from either the reactor coolant pressure boundary or the primary containment.
Page 56
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 1. Main Steam Line Isolation
: a. Reactor Vessel      6.2                Feedwater Piping Break  1) Automatic Initiation Water Level-        6.3                Feedwater Line Break        - Reactor Vessel Level - Low, Level 3 Low Low Low,        15.2.8            Outside Containment        - Reactor Vessel Water Level Low Low Level (Level 2) RCIC Level 1                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS 15.6.6
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Reactor Vessel High - Relief / LLS
                                                                              - ADS timers
: 2) Manual Containment/Drywell Isolation 6.2                Steam System Piping      1) Automatic Initiation 6.3                Break Outside              - Main Steam Line Flow - High 15.6.4            Containment                - Main Steam Isolation Valve - Closure
                                                                              - Reactor Pressure High - Relief/LLS
                                                                              - Reactor Vessel Water Level-Low Low Low, Level 1
                                                                              - ADS Initiation Timer
: 2) Manual Containment/Drywell Isolation Page 57
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 1. Main Steam Line Isolation
: a. Reactor Vessel      6.2                Loss of Coolant Accident 1) Automatic Initiation Water Level-        6.3                                            - Reactor Vessel Level - Low, Level 3 Low Low Low,        15.6.5
* Drywell pressure - High (RPS)
Level 1 - cont.                                                    - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Drywell Pressure - High (LPCI A and LPCS)
                                                                              - Drywell Pressure - High (LPCI B and LPCI C)
                                                                              - Drywell Pressure - High (HPCS)
                                                                              - Drywell Pressure - High (ADS A and E)
                                                                              - Drywell Pressure - High (ADS B and F)
: 2) Manual Containment/Drywell Isolation Page 58
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 1. Main Steam Line Isolation - cont.
: b. Main Steam Line      6.2                Pressure Regulator      1) Automatic Initiation Pressure - Low      6.2                Failure - Open              - Reactor Vessel Level - High (Level 8) Feedwater 15.1.3                                        - Reactor Water Level - High (level 8) RPS
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                              - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                              - Main Steam Line Pressure Low
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual Containment/Drywell Isolation 6.2                Recirculation Pump Trip  1) Automatic Initiation 6.3                                            - Reactor Vessel Level - High, (Level 8) Feedwater 15.3.1
* Reactor Water Level - High (level 8) RPS
                                                                              - Turbine Stop Valve Closure
* Reactor Water Level - High (level 8) RPS
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                              - Main Steam Line Pressure Low
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Main Steam Isolation Valve - Closure
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual Containment/Drywell Isolation Page 59
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 1. Main Steam Line Isolation - cont.
: b. Main Steam Line      6.2                Recirculation Pump      1) Automatic Initiation Pressure - Low -    6.3                Shaft Break                - Reactor Water Level - High (level 8) Feedwater cont.                15.3.4
* Reactor Water Level - High (level 8) RPS
                                                                              - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                              - Reactor Vessel High - Relief / LLS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                              - Main Steam Line Pressure Low
: 2) Manual Containment/Drywell Isolation 6.2                Steam System Pipe        1) Automatic Initiation 6.3                Break Outside              - Main Steam Line Flow - High 15.6.4            Containment                - Main Steam Isolation Valve - Closure
                                                                              - Reactor Pressure High - Relief/LLS
                                                                              - Reactor Vessel Water Level-Low Low Low, Level 1
                                                                              - ADS Initiation Timer
: 2) Manual Containment/Drywell Isolation Page 60
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 1. Main Steam Line Isolation - cont.
: c. Main Steam          6.2                Loss of Coolant Accident 1) Automatic Initiation Flow - High          6.3                                            - Reactor Vessel Level - Low, Level 3 15.6.5
* Drywell pressure - High (RPS)
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Drywell Pressure - High (LPCI A and LPCS)
                                                                              - Drywell Pressure - High (LPCI B and LPCI C)
                                                                              - Drywell Pressure - High (HPCS)
                                                                              - Drywell Pressure - High (ADS A and E)
                                                                              - Drywell Pressure - High (ADS B and F)
: 2) Manual Containment/Drywell Isolation
: d. Condenser            6.2                Loss of Condenser        1) Automatic Initiation Vacuum - Low        6.3                Vacuum                      - Turbine Stop Valve Closure 15.2.5
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual Containment/Drywell Isolation Page 61
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 1. Main Steam Line Isolation - cont.
: d. Condenser            6.2                Loss of AC power        1) Automatic Initiation Vacuum - Low -      6.3                                            - Reactor Vessel Level - High, (Level 8) Feedwater cont.                15.2.6
* Reactor Water Level - High (level 8) RPS
* Turbine Stop Valve Closure
                                                                                - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                                - Main Steam Isolation Valve - Closure
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                                - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                                - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                                - Condenser Vacuum - Low
                                                                                - Reactor Pressure High - Relief/LLS
: 2) Manual Containment/Drywell Isolation
: e. Main Steam          None              None                    2 out of 4 logic Tunnel Temperature -
High
: f. Main Steam Line      None              None                    2 out of 4 logic Turbine Building Temperature-High
: g. Manual Initiation    None              None                    ---
Page 62
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 2. Primary Containment and Drywell Isolation
: a. Reactor Vessel      6.2                Pressure Regulator        1) Automatic Initiation Water Level -      6.3                Failure - Open              - Reactor Vessel Level - High (Level 8) Feedwater Low Low            15.1.3                                          - Reactor Water Level - High (level 8) RPS (Level 2)
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                              - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                              - Main Steam Line Pressure Low
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual Containment/Drywell Isolation 6.2                Main Steam Line Isolation 1) Automatic Initiation 6.3                Valve (MSLIV) - Closure      - Main Steam Isolation Valve - Closure 15.2.4
* Reactor Steam Dome Pressure
                                                                              - APRM Fixed Neutron Flux - High
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                              - Reactor Pressure High ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual Containment/Drywell Isolation Page 63
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 2. Primary Containment and Drywell Isolation - cont.
: a. Reactor Vessel      6.2                Loss of Condenser        1) Automatic Initiation Water Level -      6.3                Vacuum                      - Turbine Stop Valve Closure Low Low            15.2.5
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low (Level 2) - cont
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual Containment/Drywell Isolation 6.2                Loss of AC power        1) Automatic Initiation 6.3                                            - Reactor Vessel Level - High, (Level 8) Feedwater 15.2.6
* Reactor Water Level - High (level 8) RPS
* Turbine Stop Valve Closure
                                                                              - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                              - Main Steam Isolation Valve - Closure
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Condenser Vacuum - Low
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual Containment/Drywell Isolation Page 64
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 2. Primary Containment and Drywell Isolation - cont.
: a. Reactor Vessel      6.2                Loss of Feedwater Flow  1) Automatic Initiation Water Level -      6.3                                            - Reactor Vessel Water Level Low Level (Level 3)
Low Low            15.2.7                                        - Reactor Vessel Water Level Low Low Level (Level 2) RCIC (Level 2) - cont                                                  - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
: 2) Manual Containment/Drywell Isolation 6.2                Feedwater Line Break    1) Automatic Initiation 6.3                Outside Containment        - Reactor Vessel Level - Low, Level 3 15.2.8                                          - Reactor Vessel Water Level Low Low Level (Level 2) RCIC 15.6.6                                          - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Reactor Vessel High - Relief / LLS
                                                                              - ADS timers
: 2) Manual Containment/Drywell Isolation Page 65
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 2. Primary Containment and Drywell Isolation - cont.
: a. Reactor Vessel      15.3.1            Recirculation Pump Trip  1) Automatic Initiation Water Level -                                                      - Reactor Vessel Level - High, (Level 8) Feedwater Low Low
* Reactor Water Level - High (level 8) RPS (Level 2) - cont                                                  - Turbine Stop Valve Closure
* Reactor Water Level - High (level 8) RPS
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                              - Main Steam Line Pressure Low
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Main Steam Isolation Valve - Closure
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual Containment/Drywell Isolation 15.3.3            Recirculation Pump      1) Automatic Initiation Seizure                    - Reactor Water Level - High (level 8) Feedwater
                                                                              - Reactor Water Level - High (level 8) RPS
                                                                              - Turbine Stop Valve Closure
                                                                              - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                              - Main Steam Line Pressure Low
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                              - Reactor Vessel High - Relief / LLS
: 2) Manual Containment/Drywell Isolation Page 66
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 2. Primary Containment and Drywell Isolation - cont.
: a. Reactor Vessel      15.3.4            Recirculation Pump Shaft  1) Automatic Initiation Water Level -                          Break                        - Reactor Water Level - High (level 8) Feedwater Low Low
* Reactor Water Level - High (level 8) RPS (Level 2) - cont                                                    - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                              - Reactor Vessel High - Relief / LLS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                              - Main Steam Line Pressure Low
: 2) Manual Containment/Drywell Isolation 6.2                Loss of Coolant Accidents 1) Automatic Initiation 6.3                                              - Reactor Vessel Level - Low, Level 3 15.6.5
* Drywell pressure - High (RPS)
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Drywell Pressure - High (LPCI A and LPCS)
                                                                              - Drywell Pressure - High (LPCI B and LPCI C)
                                                                              - Drywell Pressure - High (HPCS)
                                                                              - Drywell Pressure - High (ADS A and E)
                                                                              - Drywell Pressure - High (ADS B and F)
: 2) Manual Containment/Drywell Isolation Page 67
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 2. Primary Containment and Drywell Isolation - cont.
: b. Drywell Pressure    6.2                Loss of Coolant Accident 1) Automatic Initiation water level or DW press
          - High              6.3                                            - Reactor Vessel Level - Low, Level 3 15.6.5
* Drywell pressure - High (RPS)
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Drywell Pressure - High (LPCI A and LPCS)
                                                                              - Drywell Pressure - High (LPCI B and LPCI C)
                                                                              - Drywell Pressure - High (HPCS)
                                                                              - Drywell Pressure - High (ADS A and E)
                                                                              - Drywell Pressure - High (ADS B and F)
: 2) Manual Containment/Drywell Isolation
: c. deleted            None              None                    None
: d. Drywell Pressure-  6.2                Loss of Coolant Accident Same as TS 3.3.6.1.2.a High (ECCS          6.3 Divisions 1 and 2)  15.6.5
: e. Reactor Vessel      6.2                Loss of Coolant Accident Same as TS 3.3.6.1.2.a Water Level-Low    6.3 Low, Level 2        15.6.5 (HPCS NSPS Div 3 and 4)
Page 68
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 2. Primary Containment and Drywell Isolation - cont.
: f. Drywell Pressure-  6.2                Loss of Coolant Accident Same as 3.3.6.1.2.a High (HPCS NSPS    6.3 Div 3 and 4)      15.6.5
: g. Containment        6.2                Loss of Coolant Accident Same as 3.3.6.1.2.a Building Fuel      6.3 Transfer Pool      15.6.5 Ventilation Plenum Radiation-High
: h. Containment        6.2                Loss of Coolant Accident Same as 3.3.6.1.2.a Building Exhaust    6.3 Radiation-High    15.6.5
: i. Containment        6.2                Loss of Coolant Accident Same as 3.3.6.1.2.a Building            6.3 Continuous        15.6.5 Containment Purge (CCP)
Exhaust Radiation-High
: j. Reactor Vessel      6.2                Feedwater Piping Break  1) Automatic Initiation Water Level-Low    6.3                Feedwater Line Break        - Reactor Vessel Level - Low, Level 3 Low Low, Level 1  15.2.8              Outside Containment        - Reactor Vessel Water Level Low Low Level (Level 2) RCIC 15.6.6                                          - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Reactor Vessel High - Relief / LLS
                                                                              - ADS timers
: 2) Manual Containment/Drywell Isolation Page 69
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                        Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 2. Primary Containment and Drywell Isolation - cont.
: j. Reactor Vessel      6.2                Steam System Piping      1) Automatic Initiation Water Level-Low    6.3                Break Outside              - Main Steam Line Flow - High Low Low, Level 1 - 15.6.4              Containment                - Main Steam Isolation Valve - Closure cont.                                                              - Reactor Pressure High - Relief/LLS
                                                                              - Reactor Vessel Water Level-Low Low Low, Level 1
                                                                              - ADS Initiation Timer
: 2) Manual Containment/Drywell Isolation 6.2                Loss of Coolant Accident 1) Automatic Initiation 6.3                                            - Reactor Vessel Level - Low, Level 3 15.6.5
* Drywell pressure - High (RPS)
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Drywell Pressure - High (LPCI A and LPCS)
                                                                              - Drywell Pressure - High (LPCI B and LPCI C)
                                                                              - Drywell Pressure - High (HPCS)
                                                                              - Drywell Pressure - High (ADS A and E)
                                                                              - Drywell Pressure - High (ADS B and F)
: 2) Manual Containment/Drywell Isolation
: k. Containment        None              None                        ---
Pressure - High
: l. Manual Initiation  None              None                        ---
Page 70
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function          USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 3. Reactor Core Isolation Cooling (RCIC) System Isolation
: a. Auxiliary Building    None              None                      1) Automatic Isolation RCIC Steam Line                                                      - RCIC Steam Flow High + RCIC Steam Line Flow Timer Flow-High                                                        2) Manual RCIC Isolation
: b. RCIC Steam Line      None              None                      1) Automatic Isolation Flow-High, Time                                                      - RCIC Steam Flow High + RCIC Steam Line Flow Timer Delay                                                            2) Manual RCIC Isolation
: c. RCIC Steam            None              None                      1) Automatic Isolation Supply Line                                                          - RCIC Steam Supply Pressure Low Pressure-Low                                                      2) Manual RCIC Isolation
: d. RCIC Turbine          None              None                      1) Automatic Isolation Exhaust                                                              - RCIC Turbine Exhaust Diaphragm Pressure High Diaphragm                                                        2) Manual RCIC Isolation Pressure - High
: e. RCIC Equipment        None              None                      1) Automatic Initiation Room Ambient                                                        - RCIC Area Room Ambient Temperature - High Temperature -                                                    2) Manual RCIC Isolation High
: f. Main Steam Line      None              None                      1) Automatic Initiation Tunnel Ambient                                                      - Main Steam Line Tunnel Ambient Temperature-High + Main Temperature-High                                                      Steam Line Tunnel Temperature Timer
: 2) Manual RCIC Isolation
: g. Main Steam Line      None              None                      1) Automatic Initiation Tunnel                                                              - Main Steam Line Tunnel Ambient Temperature-High + Main Temperature                                                          Steam Line Tunnel Temperature Timer Timer                                                            2) Manual RCIC Isolation Page 71
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument                Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function          USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 3. Reactor Core Isolation Cooling (RCIC) System Isolation - cont.
: h. Reactor Vessel        6.2                Pressure Regulator      Same as 3.3.6.1.2.a Water Level-Low      6.3                Failure - Open Low, Level 2          15.1.3 6.2                MSLIV Closure            Same as 3.3.6.1.2.a 6.3 15.2.4 6.2                Loss of Condenser        Same as 3.3.6.1.2.a 6.3                Vacuum 15.2.5 6.2                Loss of AC Power        Same as 3.3.6.1.2.a 6.3 15.2.6 6.2                Loss of Feedwater Flow  Same as 3.3.6.1.2.a 6.3 15.2.7 6.2                Feedwater Piping Break  Same as 3.3.6.1.2.a 6.3                Feedwater Line Break 15.2.8            Outside Containment 15.6.6 6.2                Recirculation Pump Trip  Same as 3.3.6.1.2.a 6.3 15.3.1 6.2                Recirculation Pump      Same as 3.3.6.1.2.a 6.3                Seizure 15.3.3 6.2                Recirculation Pump      Same as 3.3.6.1.2.a 6.3                Shaft Break 15.3.4 6.2                Loss of Coolant Accident Same as 3.3.6.1.2.a 6.3 15.6.5 Page 72
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 3. Reactor Core Isolation Cooling (RCIC) System Isolation - cont.
: i. Drywell RCIC          None              None                    1) Bounded by Reactor Recirculation and Main Steam Line breaks Steam Line Flow -
High
: j. Drywell Pressure      6.2                Loss of Coolant Accident See 3.3.6.1.2.a
          - High                6.3 15.6.5
: k. Manual Initiation    None              None                    1) Manual Initiation
: 4. Reactor Water Cleanup (RWCU) System Isolation
: a. Differential Flow -  None              None                    1) Bounded by Reactor Recirculation and Main Steam Line breaks High
: b. Differential Flow -  None              None                    1) Bounded by Reactor Recirculation and Main Steam Line breaks Timer
: c. RWCU Heat            None              None                    1) Bounded by Reactor Recirculation and Main Steam Line breaks Exchanger Equipment Room Temperature -
High Page 73
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 4. Reactor Water Cleanup (RWCU) System Isolation - cont.
: d. RWCU Pump            None              None                    1) Bounded by Reactor Recirculation and Main Steam Line breaks Rooms Temperature-High
: e. Main Steam Line      None              None                    1) Bounded by Reactor Recirculation and Main Steam Line breaks Tunnel Ambient Temperature -
High
: f. Reactor Vessel      None              None                    1) Bounded by Reactor Recirculation and Main Steam Line breaks Water Level-Low Low, Level 2
: g. Standby Liquid      15.8.1            1) Anticipated Transient 1) Automatic Initiation Control System                            Without SCRAM            - Manual Standby Liquid Control System Initiation Initiation                                                      2) Manual RWCU Isolation
: h. Manual Initiation    None              None                    ---
3.3.6.1 5. RHR System Isolation
: a. RHR Heat            None              None                    1) Bounded by Reactor Recirculation and Main Steam Line breaks Exchanger Ambient Temperature-High Page 74
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Primary Containment Isolation Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.1 5. RHR System Isolation - cont.
: b. Reactor Vessel      None              None                    1) Bounded by Reactor Recirculation and Main Steam Line breaks.
Water Level - Low (Level 3)
: c. Reactor Vessel      None              None                    1) Bounded by Reactor Recirculation and Main Steam Line breaks.
Water Level - Low, Level 3
: d. Reactor Vessel      6.3                Feedwater Piping Break  1) Same as 3.3.6.1.2.j Water Level - Low  15.2.8            Feedwater Line Break Low Low, Level 1    15.6.6 6.3                Steam System Piping      1) Same as 3.3.6.1.2.j 15.6.4            Break 6.3                Loss of Coolant Accident 1) Same as 3.3.6.1.2.j 15.6.5
: e. Reactor Vessel      None              None                    This RHR interlock is not assumed in the accident or transient Pressure-High                                                  analysis in the USAR.
: f. Drywell Pressure-  None              None                    1) Automatic Isolation High                                                              - Drywell Pressure-High RHR High Drywell        2) Manual RHR Isolation Pressure is not modeled in any USAR accident or transient analysis because other leakage paths (e.g., MSIVs) are more limiting.
: g. Manual Initiation  None              None                    1) Manual Initiation Page 75
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
: 6.      Residual Heat Removal (RHR) Containment Spray System Instrumentation
 
==Reference:==
TS LCO 3.3.6.3 and B3.3.6.3 Residual Heat Removal (RHR) Containment Spray System Instrumentation 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 loss of coolant accident (LOCA).
The RHR Containment Spray System is automatically initiated by Reactor Vessel Water Level-Low Low Low, Level 1, Drywell Pressure-High, and Containment Pressure-High signals. The channels provide inputs to two trip systems; one trip system initiates one containment spray subsystem while the second trip system initiates the other containment spray subsystem as described in USAR section 7.3.1.1.4.
For a trip system to initiate the associated subsystem, it must receive one signal from each of the following inputs: Drywell Pressure-High, Containment Pressure-High, and a System Timer.
The Drywell Pressure-High and Containment Pressure-High Functions each have two channels, which are arranged in a one-out-of-two logic to provide the necessary signal.
The System Timer is initiated by a one-out-of-two taken twice logic consisting of two channels each of the Reactor Vessel Water Level-Low Low Low, Level 1 and Drywell Pressure-High Functions.
When the System Timer has timed out, the trip system receives the System Timer signal.
Manual initiation of the system is accomplished with the use of manual initiation push buttons.
The system can be manually initiated using the manual initiation push buttons only if a Drywell Pressure-High signal is present. There is no time delay when using the manual initiation push buttons.
Page 76
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Residual Heat Removal (RHR) Containment Spray System Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.3 Residual Heat Removal (RHR) Containment Spray System Instrumentation
: 1. Drywell Pressure  6.2                Loss of Coolant Accident 1) Automatic Initiation
          - High            6.3                                            - Reactor Vessel Level - Low, Level 3 15.6.5
* Drywell pressure - High (RPS)
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                            - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                            - Drywell Pressure - High (LPCI A and LPCS)
                                                                            - Drywell Pressure - High (LPCI B and LPCI C)
                                                                            - Drywell Pressure - High (HPCS)
                                                                            - Drywell Pressure - High (ADS A and E)
                                                                            - Drywell Pressure - High (ADS B and F)
: 2) Manual RHR containment Spray initiation
: 2. Containment        6.2                Loss of Coolant Accident 1) same as 3.3.6.3.1 above Pressure - High    6.3 15.6.5 Page 77
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Residual Heat Removal (RHR) Containment Spray System Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                        Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.3 3. Reactor Vessel    6.2                Feedwater Line Break    1) Automatic Initiation Water Level -      6.3                Outside Containment          - Reactor Vessel Level - Low, Level 3 Low Low Low        15.2.8                                          - Reactor Vessel Water Level Low Low Level (Level 2) RCIC Level 1                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS 15.6.6
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS A and E)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (ADS B and F)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI A and LPCS)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (LPCI B and LPCI C)
                                                                              - Reactor Vessel Level - Low, Low, Low Level 1 (NS4)
                                                                              - Reactor Vessel High - Relief / LLS
                                                                              - ADS timers
: 2) Manual ECCS initiation 6.2                Steam System Piping      1) Automatic Initiation 6.3                Break Outside                - Main Steam Line Flow - High 15.6.4            Containment                  - Main Steam Isolation Valve - Closure
                                                                              - Reactor Pressure High - Relief/LLS
                                                                              - Reactor Vessel Water Level-Low Low Low, Level 1
                                                                              - ADS Initiation Timer
: 2) Manual ECCS initiation 6.2                Loss of Coolant Accident Same as 3.3.6.3.1 above 6.3 15.6.5
: 4. Timers, System    None              None                    Implicit to RHR function A and System B
: 5. Timer, System B    None              None                    Implicit to RHR function only
: 6. Manual Initiation  None              None                    1) Manual Initiation coincident with High Drywell pressure Page 78
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
: 7.      Suppression Pool Makeup (SPMU) System Instrumentation
 
==Reference:==
TS LCO 3.3.6.4 and B3.3.6.4 Suppression Pool Makeup (SPMU) System Instrumentation The SPMU System provides water from the upper containment pool to the suppression pool, by gravity flow, after a loss of coolant accident (LOCA) to ensure that primary containment temperature and pressure design limits are met.
The SPMU System is automatically initiated by signals generated by:
* Reactor Vessel Water LevelLow Low Low, Level 1;
* Drywell PressureHigh; and
* Suppression Pool Water LevelLow Low channels.
The channels provide inputs to two trip systems; one trip system initiates one SPMU subsystem while the second trip system initiates the other SPMU subsystem per USAR, Section 7.3.1.1.10.
Two separate initiation logics are provided for each trip system.
The LOCA signal is received from the associated division of low pressure Emergency Core Cooling Systems (ECCS) initiation signal (i.e., two channels of Reactor Vessel Water Level Low Low Low, Level 1 and two channels of Drywell PressureHigh) are arranged in a one-out-of-two taken twice logic. The associated low pressure ECCS division's Manual Initiation push button (one per division) also supplies a signal, which manually performs the same function as the automatic LOCA signal.
Two channels of Suppression Pool Water LevelLow Low are arranged in a one-out-of-two logic, which generates the Suppression Pool Water LevelLow Low signal.
An ECCS Automatic initiation on Reactor Vessel Water Level Low Low Low Level 1 or Drywell Pressure - High or an ECCS manual initiation coincident with Suppression Pool Water Level -
Low Low will initiate the SPMU System.
Additionally, the SPMU system will initiate upon receipt of the ECCS Automatic/Manual Initiation signal after a time delay of approximately 30 minutes to ensure adequate pool level even if the Suppression Pool Low Low Signal has not yet been received.
Each SPMU dump valve (two in series in each dump line) can be manually opened from the main control room to manually initiate an associated SPMU subsystem.
The SPMU System is relied upon to dump upper containment pool water to the suppression pool to maintain drywell and horizontal vent coverage and an adequate suppression pool heat sink volume to ensure that the primary containment internal pressure and temperature stay within design limits per USAR 6.2.7, Suppression Pool Makeup System.
Page 79
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Suppression Pool Makeup (SPMU) System Instrumentation Redundancy/Diversity Instrument            Credited Safety Analysis Event                    Redundant/Diverse Instrumentation TS Function        USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.4 Suppression Pool Makeup (SPMU) System Instrumentation
: 1. Drywell            None              None                  1) Suppression pool vents are implicitly assumed to remain Pressure - High                                                covered during a LOCA in the accident analysis. One out of two logic taken twice logic.
: 2. Reactor Vessel    None              None                  1) Suppression pool vents are implicitly assumed to remain Water Level -                                                  covered during a LOCA in the accident analysis. One out of Low Low Low                                                    two logic taken twice logic.
Level 1
: 3. Suppression        None              None                  1) Suppression pool vents are implicitly assumed to remain Pool Water Level                                              covered during a LOCA in the accident analysis. One out of two
          - Low Low                                                      logic taken twice logic.
: 4. Timer              None              None                  1) Suppression pool vents are implicitly assumed to remain covered during a LOCA in the accident analysis. Two independent systems provide the function.
: 5. Manual Initiation  None              None                  ---
Page 80
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
: 8.      Relief and Low Low Set (LLS) Instrumentation
 
==Reference:==
TS LCO 3.3.6.5 and B3.3.6.5 Relief and Low Low Set (LLS) Instrumentation The safety/relief valves (S/RVs) prevent overpressurization of the nuclear steam system.
Instrumentation is provided to support two modes (in addition to the automatic depressurization system (ADS) mode of operation for selected valves) of S/RV operation:
* the relief function (all valves) and
* the LLS function (selected valves)
The relief function of the S/RVs prevents overpressurization of the nuclear steam system. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the S/RV relief function instrumentation. The LSSS establish the threshold for protective system action to prevent exceeding acceptable limits, including Safety Limits (SLs),
during Anticipated Operational Occurrences (AOOs) and Design Basis Accidents (DBAs).
The relief instrumentation consists of two trip systems, with each trip system actuating one solenoid for each S/RV. There are two solenoids per S/RV, and each solenoid can open its respective S/RV.
The relief mode (S/RVs and associated trip systems) is divided into three setpoint groups (the low with one S/RV, the medium with eight S/RVs, and the high with seven S/RVs). The outputs of the Analog Trip Modules are arranged in a two-out-of-two logic for each trip system in each setpoint group.
The LLS function of the S/RVs is designed to mitigate the effects of postulated pressure loads on the containment by preventing multiple actuations in rapid succession of the S/RVs subsequent to their initial actuation. Upon any S/RV actuation, the LLS logic assigns preset opening setpoints to two preselected S/RVs and reclosing setpoints to five preselected S/RVs.
* These setpoints are selected to override the normal relief setpoints such that the LLS S/RVs will stay open longer, thus releasing more steam (energy) to the suppression pool; hence more energy (and time) is required for repressurization and subsequent S/RV openings.
* The LLS logic is divided into three logic groups; Low, Medium and High.
o The low and medium setpoint groups each control one valve (i.e., valves 1B21F051D and 1B21-F051C, respectively).
o The high setpoint group controls the remaining three valves (i.e., valves 1B21-F047F, 1B21-F051B, and 1B21-F051G).
o The LLS logic increases the time between (or prevents) subsequent actuations to limit S/RV subsequent actuations to one valve, so that containment loads will also be reduced.
Either trip system can actuate the LLS S/RVs by energizing the associated solenoids. Each LLS trip system is enabled and sealed in upon initial S/RV actuation from the existing reactor steam dome pressure sensors of any of the normal relief setpoint groups. The reactor steam dome pressure channels that control the opening and closing of the LLS S/RVs are arranged in either a one-out-of-one or a two-out-of-two logic depending on which LLS S/RV group is being controlled.
Page 81
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Relief and Low Low Set (LLS) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function      USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.5 Relief and Low Low Set (LLS) Instrumentation
: a. Relief Mode        15.1.2            Feedwater Controller    1) Automatic Initiation Failure - Maximum          - Reactor Vessel Level - High (Level 8) Feedwater Demand                    - Reactor Water Level - High (level 8) RPS
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                            - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual SRV operation 15.1.3            Pressure Regulator      1) Automatic Initiation Failure - Open            - Reactor Vessel Level - High (Level 8) Feedwater
                                                                            - Reactor Water Level - High (level 8) RPS
* APRM Fixed Neutron Flux - High
* APRM Flow Biased Simulated Thermal Power - Upscale
                                                                            - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Main Steam Line Pressure Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual SRV operation Page 82
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Relief and Low Low Set (LLS) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                        Redundant/Diverse Instrumentation TS Function      USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.5 Relief and Low Low Set (LLS) Instrumentation
: a. Relief Mode -      15.2.2              Generator Load            1) Automatic Initiation cont.                                  Rejection                    - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
* Turbine Stop Valve Closure
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual SRV operation 15.2.3              Turbine Trip              1) Automatic Initiation
                                                                              - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual SRV operation 6.3                Main Steam Line          1) Automatic Initiation 15.2.4            Isolation Valve (MSLIV) -    - Main Steam Isolation Valve - Closure Closure
* Reactor Steam Dome Pressure
                                                                              - APRM Fixed Neutron Flux - High
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                              - Reactor Pressure High ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual SRV operation Page 83
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Relief and Low Low Set (LLS) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function      USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.5 Relief and Low Low Set (LLS) Instrumentation
: a. Relief mode -      15.2.6            Loss of AC power        1) Automatic Initiation cont.                                                            - Reactor Vessel Level - High, (Level 8) Feedwater
* Reactor Water Level - High (level 8) RPS
* Turbine Stop Valve Closure
                                                                            - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Main Steam Isolation Valve - Closure
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                            - Condenser Vacuum - Low
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual SRV operation 15.3.1            Recirculation Pump Trip 1) Automatic Initiation
                                                                            - Reactor Vessel Level - High, (Level 8) Feedwater
* Reactor Water Level - High (level 8) RPS
                                                                            - Turbine Stop Valve Closure
* Reactor Water Level - High (level 8) RPS
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Main Steam Line Pressure Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Main Steam Isolation Valve - Closure
                                                                            - Reactor Pressure High - Relief/LLS
: 2) Manual SRV operation Page 84
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Relief and Low Low Set (LLS) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                      Redundant/Diverse Instrumentation TS Function      USAR Section      Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.5 Relief and Low Low Set (LLS) Instrumentation
: a. Relief mode -      15.3.3            Recirculation Pump      1) Automatic Initiation cont.                                Seizure                    - Reactor Water Level - High (level 8) Feedwater
                                                                            - Reactor Water Level - High (level 8) RPS
                                                                            - Turbine Stop Valve Closure
                                                                            - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Main Steam Line Pressure Low
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                            - Reactor Vessel High - Relief / LLS
: 2) Manual SRV operation 15.3.4            Recirculation Pump      1) Automatic Initiation Shaft Break                - Reactor Water Level - High (level 8) Feedwater
* Reactor Water Level - High (level 8) RPS
                                                                            - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
                                                                            - Reactor Vessel High - Relief / LLS
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                            - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                            - Main Steam Line Pressure Low
: 2) Manual SRV operation Page 85
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity Relief and Low Low Set (LLS) Instrumentation Redundancy/Diversity Instrument              Credited Safety Analysis Event                        Redundant/Diverse Instrumentation TS Function      USAR Section        Transient / Accident
* diverse instrumentation not credited in the safety analysis event 3.3.6.5 Relief and Low Low Set (LLS) Instrumentation
: a. Relief mode -      6.3                Steam System Piping      1) Automatic Initiation cont.              15.6.4            Break Outside                - Main Steam Line Flow - High Containment                  - Main Steam Isolation Valve - Closure
                                                                              - Reactor Pressure High - Relief/LLS
                                                                              - Reactor Vessel Water Level-Low Low Low, Level 1
                                                                              - ADS Initiation Timer
: 2) Manual SRV operation
: b. LLS Mode          15.2.2              Generator Load            1) Automatic Initiation Rejection                    - Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
* Turbine Stop Valve Closure
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual SRV operation 15.2.3              Turbine Trip              1) Automatic Initiation
                                                                              - Turbine Stop Valve Closure
* Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
* Reactor Vessel Steam Dome Pressure High
* APRM Fixed Neutron Flux - High
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual SRV operation 6.3                  Main Steam Line          1) Automatic Initiation 15.2.4              Isolation Valve (MSLIV) -    - Main Steam Isolation Valve - Closure Closure
* Reactor Steam Dome Pressure
                                                                              - APRM Fixed Neutron Flux - High
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) RCIC
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) HPCS
                                                                              - Reactor Vessel Water Level Low Low Level (Level 2) ATWS RPT
                                                                              - Reactor Pressure High ATWS RPT
* Reactor Vessel Water Level Low Low Level (Level 2) ATWS Alternate Rod Insertion (ARI)
                                                                              - Reactor Pressure High - Relief/LLS
: 2) Manual SRV operation Page 86
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
: 9.      Loss-of-Power (LOP) Instrumentation
 
==Reference:==
TS LCO 3.3.8.1 and B3.3.8.1 Loss-of-Power (LOP) Instrumentation The LOP design creates defense-in-depth from the redundancy of the channels for the Initiation Function.
* A failed channel does not cause or prevent an initiation.
The LOP instrumentation monitors the 4.16 kV emergency buses. Offsite power is the preferred source of power for the 4.16 kV emergency buses. If the monitors determine that insufficient power is available, the buses are disconnected from the offsite power sources and connected to the onsite diesel generator (DG) power sources.
Each 4.16 kV emergency bus has its own independent LOP instrumentation and associated trip logic. The voltage for Division 1, 2, and 3, 4.16 kV buses is monitored at two levels which can be considered as two different undervoltage functions: loss of voltage and degraded voltage.
Each Division 1 and 2 emergency bus Loss of Voltage Function is monitored by two undervoltage relays on the emergency bus and two undervoltage relays on each of the two offsite power sources. The outputs of these relays are arranged in a two-out-of-two taken three times logic configuration. Each of these relays is an inverse time delay relay.
The Division 3 emergency bus Loss of Voltage Function is monitored by four undervoltage relays whose outputs are arranged in a one-out-of-two taken twice logic configuration. The output of this logic inputs to a time delay relay.
Each Division 1, Division 2, and Division 3 emergency bus Degraded Voltage Function is monitored by two undervoltage relays for each emergency bus whose outputs are arranged in a two-out-of-two logic configuration. The output of this logic inputs to a time delay relay or each emergency bus.
Accident analyses credit the loading of at least two of the DGs based on the loss-of-offsite power coincident with a loss of coolant accident (LOCA).
* Six channels for Division 1 and 2 - 4.16kV Emergency Bus Undervoltage (Loss of Voltage) Function per associated emergency bus are 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.
* Four channels For Division 3 - 4.16kV Emergency Bus Undervoltage (Loss of Voltage)
Function per associated emergency bus are 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.
* Two channels of each 4.16kV Emergency Bus Undervoltage (Degraded Voltage)
Function per associated emergency bus are 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.
Page 87
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity 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, Section 2.1.1 - Defense-in-Depth Defense-in-depth consists of several elements and consistency with the defense-in-depth philosophy is maintained if the following occurs:
* A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation.
o Current Technical Specifications (TS) reflect this balance by allowing one sensor module or channel to be placed in trip, while preserving the fundamental safety function of the applicable system. Tripping an inoperable channel does not affect the number of channels required to provide the safety function. Even in the TS condition for two channels in a function inoperable, the fundamental safety function is preserved since sufficient operable channels remain in the function.
* Over-reliance on programmatic activities as compensatory measures associated with the change in the licensing basis is avoided.
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.
* 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 (diversity), the number of minimum operable channels to perform the safety function has not been decreased, and the channels remain independent as originally designed, even with one channel inoperable.
* Defenses against potential common-cause failures are preserved, and the potential for the introduction of new common-cause failure mechanisms is assessed.
o This LAR does not impact the original determination of common-cause failure for the applicable instrumentation and its functions. It may allow the allowed outage time to be extended for one or two channels in a function to be inoperable prior to placing the channel in trip. Placing the channel in trip fulfils one of the two required channels in trip needed to perform the safety function.
* Independence of barriers is not degraded.
o Barriers are not affected by this LAR request.
Page 88
 
ATTACHMENT 5 Information Supporting Instrumentation Redundancy and Diversity
* Defenses against human errors are preserved.
o In the conditions listed in the TS, a potential extension of the allowed outage time does not change any personnel actions required when the TS Action is entered.
Therefore, no change to the possibility of a human error is introduced and no change to the defenses against that potential human error have been altered.
* The intent of the plant's design criteria is maintained.
o The design criteria of the applicable systems are maintained as reflected in the Updated Safety Analysis Report (USAR). Redundancy, diversity of signal and independence of trip channel functions are maintained with the requested change.
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 Risk Informed Completion Time (RICT) Program, to perform actions that the NRC has previously determined to be acceptable.
Therefore, the defense-in-depth principals prescribed in Regulatory Guide 1.174, Revision 2, are met.
Page 89
 
ENCLOSURE 1 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" List of Revised Required Actions to Corresponding PRA Functions
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions
: 1. Introduction Section 4.0, Item 2 of the Nuclear Regulatory Commission's (NRC) Final Safety Evaluation (Reference 1) for NEI 06-09, Revision 0-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, (Reference 2) identifies the following needed content:
* The License Amendment Request (LAR) will provide identification of the Time Sensitive (TS) Limiting Conditions for Operation (LCOs) and action requirements to which the RMTS will apply.
* The LAR will provide a comparison of the TS functions to the PRA modeled functions of the Structures, Systems, and Components (SSCs) subject to those LCO actions.
* The comparison should justify that the scope of the PRA model, including applicable success criteria such as number of SSCs required, flow rate, etc., are consistent with licensing basis assumptions (i.e., 50.46 ECCS flowrates) for each of the TS requirements, or an appropriate disposition or programmatic restriction will be provided.
This enclosure provides confirmation that the Clinton Power Station (CPS) 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. The CPS PRA model has the capability to model directly or through use of a bounding surrogate the risk impact of entering each of the TS LCOs in the scope of the RICT Program.
Table E1-1 below lists each TS LCO Condition to which the RICT Program is proposed to be applied and documents the following information regarding the TSs with the associated safety analyses, the analogous PRA functions and the results of the comparison:
    -  Column "Tech Spec Description": Lists the LCOs and condition statements within the scope of the RICT Program.
    -  Column "SSCs Covered by TS LCO Condition": The SSCs addressed by each action requirement.
    -  Column "Modeled in PRA": Indicates whether the SSCs addressed by the TS LCO Condition are included in the PRA.
    -  Column "Function Covered by TS LCO Condition": A summary of the required functions from the design basis analyses.
    -  Column "Design Success Criteria": A summary of the success criteria from the design basis analyses.
    -  Column "PRA Success Criteria": The function success criteria modeled in the PRA.
    -  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 Revision 0-A.
E1-1
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions 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 (CRMP) tool for the RICT program. Differences in success criteria typically arise due to the requirement in the PRA standard to make PRAs realistic rather than bounding, whereas design basis criteria are necessarily conservative and bounding. The use of realistic success criteria is necessary to conform to capability Category II of the PRA standard as required by NEI 06-09 Revision 0-A.
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 technical specifications modeled in the PRA. Following 4b implementation, 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, Revision 0-A and the NRC safety evaluation, and may differ from the RICTs presented.
Table E1-3 lists the TSTF-505 Rev 2 Table 1 Tech Specs that require additional justification along with a description of how the additional justification is provided in the LAR.
E1-2
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered      Design CPS Tech  Tech Spec                            Modeled                                            PRA Success by TS LCO                    by TS LCO            Success                            Comments Spec      Description                          in PRA                                            Criteria Condition                    Condition            Criteria SSCs are modeled consistent with the TS scope and so can One SLC                                                                                                  be directly included SLC injection                          Same as Design 3.1.7.A  subsystem          SLC Trains        Yes                              One of two trains                  in the RTR tool for capability                              Success Criteria inoperable.                                                                                              the RICT program.
The success criteria are consistent with the design basis.
Individual RPS instrumentation inputs to the RPS logic system are not RPS                                                                                                      modeled in the PRA.
Instrumentation -                                                                                        A surrogate is Four reactor                  Provide reactor trip  2 of 4 channels One or more                          Not                                                Same as Design  chosen and it 3.3.1.1.A                  protection                    signal based on plant Functions with                      explicitly                                        Success Criteria represents the system channels              parameters            See Note 1.
one channel                                                                                              common cause inoperable.                                                                                              failure of the RPS electrical system.
This is conservative as it represents failure of the RPS.
RPS Instrumentation -
One or more 3.3.1.1.B                  See 3.3.1.1.A Functions with two channels inoperable.
E1-3
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered      Design CPS Tech  Tech Spec                            Modeled                                            PRA Success by TS LCO                    by TS LCO              Success                          Comments Spec      Description                          in PRA                                            Criteria Condition                    Condition              Criteria End of Cycle Recirculation                                                                                            Not explicitly Pump Trip                                                                                                modeled in the PRA.
(EOC-RPT)                                      Minimize reactor                                        Conservatively this Recirculation                                        2 of 4 channels Instrumentation -                    Not        pressure increase by                    Same as Design  would be treated as 3.3.4.1.A                  pump One or more                          explicitly reduction in core flow                  Success Criteria equivalent to the instrumentation                                      See Note 2.
Functions with                                  after reactor trip                                      ATWS-RPT and one required                                                                                            assessed as such in channel                                                                                                  the RICT calculation.
inoperable.
End of Cycle Recirculation Pump Trip (EOC-RPT) 3.3.4.1.B Instrumentation - See 3.3.4.1.A One or more Functions with two channels inoperable.
One or more required                                                                                                ECCS channels                                                                                                instrumentation is inoperable as                                                                                            not explicitly required by                                                                                              modeled. Therefore, 1 out of 2 taken Action A.1 and                                                                                          the frontline system ECCS actuation    Not        ECCS initiation and    twice            Same as Design 3.3.5.1.B referenced in                                                                                            is mapped as a instrumentation    explicitly logic instrumentation                  Success Criteria Table 3.3.5.1                                                                                        surrogate. The See Note 3.
Reactor low-low-                                                                                        success criteria are low level,                                                                                              consistent with the Reactor Shroud                                                                                          design basis. See level and Drywell                                                                                        Note 3.
high pressure E1-4
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered      Design CPS Tech  Tech Spec                            Modeled                                          PRA Success by TS LCO                    by TS LCO            Success                          Comments Spec      Description                          in PRA                                            Criteria Condition                    Condition            Criteria One or more                                                          1 out of 2 taken required                                                              twice for low channels                                                              pressure                          ECCS inoperable as                                                        injection                        instrumentation is required by                                                          permissive                        not explicitly Action A.1 and                                                                                          modeled. Therefore, referenced in                                                        Two time delays                  the frontline system ECCS actuation    Not        ECCS initiation and                    Same as Design 3.3.5.1.C Table 3.3.5.1-1                                                      associated with                  is mapped as a instrumentation    explicitly logic instrumentation                  Success Criteria
          - Reactor low                                                        LPCI A and                        surrogate. The pressure                                                              LPCI B. Loss of                  success criteria are injection                                                            one timer does                    consistent with the permissives and                                                      not preclude                      design basis. See LPCI / LPCS                                                          ECCS initiation.                  Note 3.
injection time delays                                                                See Note 3.
One or more ECCS required instrumentation is channels not explicitly inoperable as modeled. Therefore, required by 1 out of 2                        the frontline system Action A.1 and    ECCS actuation    Not        ECCS initiation and                    Same as Design 3.3.5.1.D                                                                                                        is mapped as a referenced in    instrumentation    explicitly logic instrumentation                  Success Criteria See Note 3.                      surrogate. The Table 3.3.5.1                                                                                                                  success criteria are RCIC Storage consistent with the Tank low level design basis. See and Suppression Note 3.
Pool high level E1-5
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered      Design CPS Tech  Tech Spec                            Modeled                                          PRA Success by TS LCO                    by TS LCO            Success                          Comments Spec      Description                          in PRA                                            Criteria Condition                    Condition            Criteria One or more required ECCS channels Remaining                        instrumentation is inoperable as capability of                    not explicitly required by ECCS                              modeled. Therefore, Action A.1 and subsystems and                    the frontline system referenced in    ECCS actuation    Not        ECCS initiation and                    Same as Design 3.3.5.1.E                                                                      redundancy                        is mapped as a Table 3.3.5.1 instrumentation    explicitly logic instrumentation                  Success Criteria available in the                  surrogate. The HPCS high ECCS design.                      success criteria are discharge consistent with the pressure and See Note 3.                      design basis. See LPCI / LPCS /
Note 3.
HPCS low flow bypass One or more required channels                                                                                                The ADS system inoperable as                                                                                          instrumentation is required by                                                                                            not modeled in Action A.1 and                                                                                          detail. Therefore, a 1 out of 2 taken referenced in                                                                                          surrogate is chosen ADS                Not        ECCS initiation and  twice            Same as Design 3.3.5.1.F Table 3.3.5.1                                                                                      that represents the instrumentation    explicitly logic instrumentation                  Success Criteria ADS Reactor                                                                                            SRVs. The success See Note 3.
low- low-low                                                                                            criteria are level, Reactor                                                                                          consistent with the confirmatory low                                                                                        design basis. See level, and                                                                                              Note 4.
Drywell high pressure E1-6
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered      Design CPS Tech  Tech Spec                            Modeled                                          PRA Success by TS LCO                    by TS LCO            Success                          Comments Spec      Description                          in PRA                                            Criteria Condition                    Condition            Criteria One or more required                                                                                                The ADS system 1 out of 2 for channels                                                                                                instrumentation is discharge inoperable as                                                                                          not modeled in pressure required by                                                                                            detail. Therefore, a Action A.1 and                                                                                          surrogate is chosen ADS                Not        ECCS initiation and  Trip system 1 or Same as Design 3.3.5.1.G referenced in                                                                                          that represents the instrumentation    explicitly logic instrumentation Trip system 2    Success Criteria Table 3.3.5.1                                                                                      SRVs. The success will initiate on ADS / LPCI /                                                                                            criteria are loss of timer.
LPCS discharge                                                                                          consistent with the pressure                                                                                                design basis. See See Note 3.
permissive and                                                                                          Note 3.
timer The RCIC system RCIC System instrumentation is Instrumentation -
not modeled in As required by detail. Therefore, a Required Action                                                      1 out of 2 taken surrogate is chosen A.1 and          RCIC actuation    Not        ECCS initiation and  twice            Same as Design 3.3.5.3.B                                                                                                        that represents the referenced in    instrumentation    explicitly logic instrumentation                  Success Criteria RCIC system. The Table 3.3.5.3                                                    See Note 4.
success criteria are Reactor Vessel consistent with the Water Level-Low design basis. See Low, Level 2 Note 4.
E1-7
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered      Design CPS Tech  Tech Spec                            Modeled                                        PRA Success by TS LCO                    by TS LCO            Success                        Comments Spec      Description                          in PRA                                        Criteria Condition                    Condition            Criteria RCIC System                                                                                          The RCIC system Instrumentation -                                                                                    instrumentation is As required by                                                                                      not modeled in Required Action  RCIC storage                                                                      detail. Therefore, a A.1 and          tank and                                            1 out of 2                    surrogate is chosen Not        ECCS initiation and                Same as Design 3.3.5.3.D referenced in    suppression                                                                        that represents the explicitly logic instrumentation              Success Criteria Table 3.3.5.3 pool level                                          See Note 4.                    RCIC system. The RCIC storage      instrumentation                                                                    success criteria are tank low level                                                                                      consistent with the and suppression                                                                                      design basis. See pool high level                                                                                      Note 4.
Primary The logic for primary Containment containment isolation and Drywell is not modeled in Isolation                                                            2 out of 4 Main Steam Line    Not        Isolate Main Steam                  Same as Design  detail. Therefore, a 3.3.6.1.A Instrumentation -
isolation (MSIV)  explicitly Line                                Success Criteria surrogate is chosen One or more                                                          See Note 5.
that represents the Functions with failure of the MSIVs one channel to isolate.
inoperable.
Primary Containment and Drywell Isolation 3.3.6.1.B Instrumentation - See 3.3.6.1.A One or more Functions with two channels inoperable.
E1-8
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered    Design CPS Tech  Tech Spec                            Modeled                                          PRA Success by TS LCO                    by TS LCO            Success                          Comments Spec      Description                          in PRA                                          Criteria Condition                    Condition            Criteria The logic for primary Primary containment isolation Containment                                                          1 out of 2 twice Primary                                                                              is not modeled in and Drywell                                                          2 out of 2 twice Containment                                                                          detail. Therefore, a Isolation                                      Primary and Drywell        Not                                              Same as Design  surrogate is chosen 3.3.6.1.D Instrumentation -                              containment/drywell  Two trip systems Isolation          explicitly                                      Success Criteria that represents either One or more                                    isolation            for 1 out of 1 EXCEPT Main                                                                          a failure of required Steam Line                                                                          containment or channels                                                            See Note 5.
failure of the frontline inoperable.
system.
1 out of 2 for RHR                                                                  Drywell Pressure Containment                                                          - High and Spray System                                                        Containment Instrumentation -                                                    Pressure - High As required by                                                                                        The logic for RHR Required Action                                                      1 out of 2 twice                  Containment Spray A.1 and          RHR                                                for timer                        is not modeled in referenced in    containment        Not        RHR Containment      initiation on    Same as Design  detail. Therefore, a 3.3.6.3.B Table 3.3.6.3 spray system      explicitly Spray initiation    water level low  Success Criteria surrogate is chosen High drywell      instrumentation                                    low low level 1                  that represents a pressure, high                                                      and Drywell                      failure of the frontline containment                                                          pressure - high                  system.
pressure, and Reactor Vessel                                                      Two trip systems Water Level-Low                                                      for timer Low, Level 1 See Note 6.
E1-9
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered    Design CPS Tech  Tech Spec                            Modeled                                          PRA Success by TS LCO                    by TS LCO            Success                            Comments Spec      Description                          in PRA                                            Criteria Condition                    Condition            Criteria RHR Containment                                                          Two trip systems                  The logic for RHR Spray System                                                        each with its                      Containment Spray Instrumentation - RHR                                                own timer. One                    is not modeled in As required by    containment        Not        RHR Containment      trip system      Same as Design  detail. Therefore, a 3.3.6.3.C Required Action  spray system      explicitly Spray initiation    fulfills the      Success Criteria surrogate is chosen A.1 and          instrumentation                                    Function.                          that represents a referenced in                                                                                          failure of the frontline Table 3.3.6.3                                                    See Note 6.                        system.
Timers 1 out of 2 twice for Reactor Vessel Water Level low low low (level 1) and Upper pool dump SPMU System                                                          Drywell pressure valves used as Instrumentation -                                                    - high Provide water from                                      surrogate. The As required by Upper pool        Not        the upper                              Same as Design  SPMU system in its 3.3.6.4.B Required Action                                                      1 out of 2 for dump valves        explicitly containment pool to                    Success Criteria entirety is not A.1 and                                                              Suppression the suppression pool                                    modeled due to referenced in                                                        Pool Water limited value to the Table 3.3.6.4-1.                                                    Level low low PRA.
Two trip systems for Timer See Note 7.
E1-10
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered      Design CPS Tech  Tech Spec                            Modeled                                            PRA Success by TS LCO                      by TS LCO            Success                            Comments Spec      Description                          in PRA                                            Criteria Condition                      Condition            Criteria Upper pool dump SPMU System valves used as Instrumentation -                                                      Two trip systems Provide water from                                      surrogate. The As required by                                                        for Timer Upper pool          Not        the upper                              Same as Design  SPMU system in its 3.3.6.4.C Required Action                                                        (Function 4) dump valves        explicitly containment pool to                    Success Criteria entirety is not A.1 and the suppression pool                                    modeled due to referenced in                                                          See Note 7.
limited value to the Table 3.3.6.4-1.
PRA.
Prevent overpressurization of Relief and Low-                                  the nuclear steam                                        Relief function is not One of two trip Low Set                                          system and ensure                                        modeled. The Low-Safety/Relief      Not                              systems.          Same as Design 3.3.6.5.A Instrumentation -                                that the containment                                    Low Set valves are Valve              explicitly                                        Success Criteria One trip system                                  loads remain within                                      used as a surrogate See Note 8.
inoperable.                                      the primary                                              for the trip systems.
containment design basis 2 out of 2 taken The LOP                                              three times -
System includes                                      Loss of Voltage Individual instrument sensors, relays,                                    for Div 1 and 2 channels for loss of bypass Loss of Power                                                                                            power capability, circuit                                  1 out of 2 taken (LOP)                                                                                                    instrumentation are breakers, and                                        twice - Loss of Instrumentation -                    Not        Undervoltage sensing                    Same as Design  not modeled.
3.3.8.1.A                  switches that are                                    Voltage for Div 3 One or more                          explicitly capability                              Success Criteria Therefore, a necessary to trip channels                                                                                                  surrogate relay is offsite power                                        2 out of 2 logic inoperable.                                                                                              chosen that fails the circuits and start                                  for Degraded DG start mode or the emergency                                        Voltage Div 1, 2, undervoltage relay.
diesel                                              and 3 generators.
See Note 9.
E1-11
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered      Design CPS Tech Tech Spec                            Modeled                                          PRA Success by TS LCO                    by TS LCO              Success                            Comments Spec    Description                          in PRA                                            Criteria Condition                    Condition              Criteria SSCs are modeled One low consistent with the pressure ECCS    Three LPCI Low pressure                            One subsystem      TS scope and so can 3.5.1.A  injection/spray  trains and one      Yes                              Two subsystems injection into the RPV                  (Ref 3)            be directly included subsystem        LPCS train in the RTR tool for inoperable.
the RICT program.
SSCs are modeled consistent with the TS scope and so can High Pressure be directly included Core Spray        HPCS                          High pressure                          Same as Design 3.5.1.B                                        Yes                              One of one train                    in the RTR tool for (HPCS) System    components                    injection into the RPV                  Success Criteria the RICT program.
inoperable.
The success criteria are consistent with the design basis.
Two ECCS injection SSCs are modeled subsystems consistent with the inoperable OR LPCI, LPCS,                                                          One subsystem      TS scope and so can 3.5.1.C  One ECCS                              Yes      Injection into the RPV Two subsystems and HPCS                                                              (Ref 3)            be directly included injection and one in the RTR tool for ECCS spray the RICT program.
subsystem inoperable.
SSCs are modeled consistent with the ADS valves and One ADS valve                                  Vessel                Six of seven    Three of seven ADS TS scope and so can 3.5.1.E                    supporting          Yes inoperable.                                    depressurization      ADS valves      valves (Ref 3)    be directly included components in the RTR tool for the RICT program.
E1-12
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered      Design CPS Tech  Tech Spec                            Modeled                                          PRA Success by TS LCO                    by TS LCO            Success                          Comments Spec      Description                          in PRA                                            Criteria Condition                    Condition            Criteria One ADS valve inoperable AND One low 3.5.1.F  pressure ECCS    See LCO Conditions 3.5.1.A and 3.5.1.C injection/spray subsystem inoperable.
SSCs are modeled consistent with the TS scope and so can Supply high pressure                                    be directly included RCIC System      RCIC                                                                Same as Design 3.5.3.A                                        Yes        makeup water to the  One of one train                  in the RTR tool for inoperable.      components                                                          Success Criteria RPV.                                                    the RICT program.
The success criteria are consistent with the design basis.
One or more                                                                                            The airlocks are not required primary                                                                                        modeled so a large containment air  Primary                                            One of two                        pre-existing leak Not                                              Same as Design 3.6.1.2.C locks inoperable  containment air              Isolation            doors maintain                    failure will be used explicitly                                        Success Criteria for reasons other lock equipment                                      boundary                          as a conservative than Condition A                                                                                        surrogate for the or B.                                                                                                  RICT calculation.
Not all primary One or more                                    Minimize the loss of containment isolation penetration flow                                reactor coolant valves are modeled.
paths with one    Primary                      inventory, and        One of two Not                                              Same as Design  For valves that are 3.6.1.3.A PCIV inoperable,  Containment                  establish the primary isolation valves explicitly                                        Success Criteria not modeled, a except due to    Isolation Valves              containment          per penetration surrogate of a pre-leakage not                                    boundary during existing containment within limit.                                  accidents.
failure is chosen.
E1-13
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered        Design CPS Tech  Tech Spec                          Modeled                                              PRA Success by TS LCO                    by TS LCO              Success                            Comments Spec      Description                        in PRA                                              Criteria Condition                    Condition              Criteria One or more Not all primary penetration flow                              Minimize the loss of containment isolation paths with one                                reactor coolant valves are modeled.
or more primary  Primary                      inventory and          One of two Not                                                  Same as Design  For valves that are 3.6.1.3.D containment      Containment                  establish the primary  isolation valves explicitly                                          Success Criteria not modeled, a purge valves not Purge Valves                  containment            per penetration surrogate of a pre-within purge                                  boundary during existing containment valve leakage                                  accidents.
failure is chosen.
limits.
SSCs are modeled Prevents excessive consistent with the short duration S/RV One LLS valve                                                          4 of 5 LLS        2 of five LLS    TS scope and so can 3.6.1.6.A                  LLS S/RVs          Yes        cycles with valve inoperable.                                                            valves            valves (Ref 3)  be directly included actuation at the relief in the RTR tool for setpoint.
the RICT program.
SSCs are modeled consistent with the TS scope and so can One RHR RHR                          Depressurize and                                          be directly included containment                                                                              Same as Design 3.6.1.7.A                  containment        Yes        cool the containment    One of two trains                  in the RTR tool for spray subsystem                                                                          Success Criteria spray systems                atmosphere.                                                the RICT program.
inoperable.
The success criteria are consistent with the design basis.
SSCs are modeled consistent with the One RHR                                                                                                  TS scope and so can suppression pool RHR pumps,                                                                              be directly included Removal of heat from                      Same as Design 3.6.2.3.A cooling          valves and heat    Yes                                One of two trains                  in the RTR tool for the Suppression Pool                      Success Criteria subsystem        exchangers                                                                              the RICT program.
inoperable.                                                                                              The success criteria are consistent with the design basis.
E1-14
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                  Function Covered      Design CPS Tech  Tech Spec                          Modeled                                            PRA Success by TS LCO                    by TS LCO            Success                              Comments Spec      Description                        in PRA                                            Criteria Condition                    Condition            Criteria One SPMU Upper pool dump subsystem Maintain suppression                                        valves used as inoperable for                      Not                                                Same as Design 3.6.2.4.C                  SPMU system                  pool water level      One of two trains                    surrogate reasons other                      explicitly                                        Success Criteria within safety limits                                        representative of the than Condition A entire system.
or B.
The drywell air lock Drywell air lock                                                                                          is not modeled in inoperable for                                Primary                                Maintain primary    detail. Therefore, a Not                              One of two 3.6.5.2.C reasons other    Drywell air lock              containment/drywell                    containment/drywell surrogate is chosen explicitly                      drywell doors than Condition A                              isolation                              isolation          that represents a or B.                                                                                                      failure of containment.
The drywell isolation One or more                                                                                                valves are not penetration flow                                                                                          modeled in detail.
Primary              One of two paths with one  Drywell isolation                                                    Same as Design      Therefore, a 3.6.5.3.A                                    Yes        containment/drywell  isolation valves required drywell valves                                                                Success Criteria    surrogate is chosen isolation            per penetration isolation valve                                                                                            that represents a inoperable.                                                                                                failure of containment.
SSCs are modeled consistent with the TS scope and so can Provide cooling water Division 1 or 2                                                                                            be directly included for the removal of                      Same as Design 3.7.1.B  SX subsystem    SX system          Yes                              One of two trains                    in the RTR tool for heat from unit                          Success Criteria inoperable.                                                                                                the RICT program.
auxiliaries The success criteria are consistent with the design basis.
E1-15
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                Function Covered      Design CPS Tech Tech Spec                          Modeled                                          PRA Success by TS LCO                    by TS LCO            Success                            Comments Spec    Description                        in PRA                                          Criteria Condition                    Condition            Criteria SSCs are modeled consistent with the Main Turbine                                Control steam                                            TS scope and so can Bypass System -                              pressure when                                            be directly included Turbine bypass                                                      Same as Design 3.7.6.A  Requirements of                    Yes      reactor steam        All 6 valves                      in the RTR tool for valves                                                              Success Criteria the LCO not                                  generation exceeds                                      the RICT program.
met.                                        turbine requirements                                    The success criteria are consistent with the design basis.
The Station                                                                          SSCs are modeled Reserve                                                                              consistent with the One of two Auxiliary                                                                            TS scope and so can qualified offsite One offsite    Transformer                  Supply AC loads                                          be directly included sources to        Same as Design 3.8.1.A  circuit        (RAT), ERAT,        Yes      during normal                                            in the RTR tool for ensure            Success Criteria inoperable. and associated              operation                                                the RICT program.
availability of breakers with                                                                        The success criteria power offsite power                                                                        are consistent with supplies                                                                              the design basis.
SSCs are modeled consistent with the TS scope and so can Supply AC loads                                          be directly included One required    EDGs and their                                    One diesel per    Same as Design 3.8.1.B                                      Yes      during abnormal                                          in the RTR tool for DG inoperable. support systems                                    division          Success Criteria operation                                                the RICT program.
The success criteria are consistent with the design basis.
Two offsite 3.8.1.C  circuits        See LCO 3.8.1.A inoperable.
E1-16
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                Function Covered      Design CPS Tech Tech Spec                            Modeled                                          PRA Success by TS LCO                    by TS LCO            Success                          Comments Spec    Description                          in PRA                                          Criteria Condition                    Condition            Criteria One offsite circuit inoperable 3.8.1.D  AND One            See LCO 3.8.1.A and 3.8.1.B required DG inoperable.
SSCs are modeled consistent with the TS scope and so can One battery be directly included charger on        Battery                      Maintaining battery  One per required Same as Design 3.8.4.A                                        Yes                                                              in the RTR tool for Division 1 or 2    Chargers                    bank fully charged    division        Success Criteria the RICT program.
inoperable.
The success criteria are consistent with the design basis.
SSCs are modeled consistent with the TS scope and so can Provide both motive One battery on                                                                                          be directly included and control power to  One per required Same as Design 3.8.4.B  Division 1 or 2    DC Batteries      Yes                                                              in the RTR tool for selected safety      division        Success Criteria inoperable.                                                                                            the RICT program.
related equipment.
The success criteria are consistent with the design basis.
The Division 1 SSCs are modeled Division 1 or 2    and 2 battery,              Provides the AC consistent with the DC electrical      associated                  emergency power TS scope and so can power              battery charger,            system with control be directly included subsystem          and all the                  power and provides    One per required Same as Design 3.8.4.C                                        Yes                                                              in the RTR tool for inoperable for    associated                  both motive and      division        Success Criteria the RICT program.
reasons other      control                      control power to The success criteria than Condition A  equipment and                selected safety are consistent with or B.              interconnecting              related equipment the design basis.
cabling.
E1-17
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                Function Covered      Design CPS Tech Tech Spec                          Modeled                                          PRA Success by TS LCO                    by TS LCO            Success                            Comments Spec    Description                        in PRA                                            Criteria Condition                    Condition            Criteria SSCs are modeled consistent with the Provide AC electrical                                    TS scope and so can Division 1 or 2                              power to the                                            be directly included One inverter per  Same as Design 3.8.7.A  inverter        Inverters          Yes      uninterruptible AC                                      in the RTR tool for required division Success Criteria inoperable.                                  and RPS solenoid                                        the RICT program.
buses                                                    The success criteria are consistent with the design basis.
SSCs are modeled 4.16 kV ESF One or more                                                                                          consistent with the buses, 480 V Division 1 or 2                                                                                      TS scope and so can ESF load AC electrical                                AC power distribution                                    be directly included centers and                                                          Same as Design 3.8.9.A  power                              Yes      to the required      One subsystem                      in the RTR tool for distribution                                                        Success Criteria distribution                                Divisional Loads                                        the RICT program.
panels, and subsystems                                                                                            The success criteria distribution inoperable.                                                                                          are consistent with panels the design basis.
SSCs are modeled One or more                                                                                          consistent with the Division 1 or 2                                                                                      TS scope and so can uninterruptible 120 V                                                                                be directly included Uninterruptible AC                      Same as Design 3.8.9.B  AC bus          uninterruptable    Yes                            One subsystem                      in the RTR tool for power distribution                      Success Criteria distribution    AC buses                                                                              the RICT program.
subsystems                                                                                            The success criteria inoperable.                                                                                          are consistent with the design basis.
E1-18
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-1: In Scope TS/LCO Conditions to Corresponding PRA Functions SSCs Covered                Function Covered      Design CPS Tech Tech Spec                          Modeled                                        PRA Success by TS LCO                    by TS LCO            Success                        Comments Spec    Description                        in PRA                                        Criteria Condition                    Condition            Criteria The Division 1 SSCs are modeled and 2 battery, One or more                                                                                        consistent with the associated Division 1 or 2                                                                                    TS scope and so can battery charger, DC electrical                                DC power distribution                                be directly included and all the                                                      Same as Design 3.8.9.C  power                              Yes      to the required      One subsystem                  in the RTR tool for associated                                                        Success Criteria distribution                                Divisional Loads                                      the RICT program.
control subsystems                                                                                        The success criteria equipment and inoperable.                                                                                        are consistent with interconnecting the design basis.
cabling.
E1-19
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Notes:
: 1. The RPS (reactor protection system) is comprised of four independent trip logic divisions (1, 2, 3, and 4). Each RPS input for a variable is independently monitored by one instrument channel in each of the four divisions. Each instrument channel combines the four RPS Function inputs for that variable in a two-out-of-four logic. Each instrument channel in turn provides an input to all four RPS trip logic divisions. The four RPS trip logic divisions are also combined in a two-out-of-four arrangement. Each RPS trip logic division provides four output signals to load drivers which de-energize the scram pilot valve solenoids. RPS success is dependent on the operability of the individual instrumentation channel Functions specified in Table 3.3.1.1-1 of the CPS Tech Specs. Loss of any functional input does not prevent the channel from responding to other inputs. Use of an electrical SCRAM failure as a surrogate for a non-modeled functional input is conservative as it encompasses loss of all the inputs to all channels rather than any single input to any channel.
: 2. The EOC-RPT system is a two-out-of-four logic for each Function; thus, either two TSV Closure or two TCV Fast Closure, Trip Oil Pressure-Low signals are required to actuate tripping both recirculation pumps from fast speed operation. There are two EOC-RPT breakers in series per recirculation pump. A trip in Division 1 (or 4) will cause a trip of the 'A' recirculation pump. A trip in Division 2 (or 3) will cause a trip of the 'B' recirculation pump. Both EOC-RPT breakers for each recirculation pump trip upon actuation of the EOC-RPT system. Loss of any functional input does not prevent the channel from responding to other inputs. Use of an ATWS-RPT surrogate for a non-modeled functional input is conservative as it encompasses loss of all the inputs to all channels rather than any single input to any channel.
: 3. The ECCS instrumentation design criteria is dependent upon the individual instrumentation channel Functions specified in Table 3.3.5.1-1.
Each Function must have a required number of operable channels, with their setpoints within the specified Allowable Values, where appropriate. In the case that individual channel or relay logic is not modeled explicitly, the affected component(s) or system(s) are used as a surrogate.
: 4. The RCIC instrumentation design criteria is dependent upon the individual instrumentation channel Functions specified in Table 3.3.5.1-1 of the CPS Tech Specs. Each Function must have a required number of operable channels, with their setpoints within the specified Allowable Values, where appropriate. Since the RCIC instrumentation is not explicitly modeled, the RCIC system itself is used as a surrogate.
: 5. The success of the primary containment and drywell isolation instrumentation is dependent on the operability of the individual instrumentation channel Functions specified in Table 3.3.6.1-1 of the CPS Tech Specs. Each Function must have a required number of operable channels, with their setpoints within the specified Allowable Values, where appropriate.
: 6. The channels provide inputs to two trip systems; one trip system initiates one containment spray subsystem while the second trip system initiates the other containment spray subsystem. For a trip system to initiate the associated subsystem, it must receive one signal from each of the following inputs: Drywell Pressure-High, Containment Pressure-High, and a System Timer. The Drywell Pressure-High and Containment Pressure-High Functions each have two channels, which are arranged in a one-out-of-two logic to provide the necessary E1-20
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions signal. The System timer is initiated by the Reactor Vessel Water Level-Low Low Low, Level 1 and Drywell Pressure-High Functions in a one out of two taken twice to allow for RHR LPCI mode injection to occur before being used in the RHR Containment Spray mode if Containment Pressure - High exists.
: 7. For the SPMU instrumentation, one initiation logic for a trip system will initiate the associated subsystem if a LOCA signal coincident with a Suppression Pool Water LevelLow Low signal is received. The LOCA signal is received from the associated division of low pressure Emergency Core Cooling Systems (ECCS) initiation signal (i.e., two channels of Reactor Vessel Water LevelLow Low Low, Level 1 and two channels of Drywell PressureHigh are arranged in a one-out-of-two taken twice logic). Two channels of Suppression Pool Water LevelLow Low are arranged in a one-out-of-two logic, which generates the Suppression Pool Water LevelLow Low signal. The associated low pressure ECCS division's Manual Initiation push button (one per division) also supplies a signal, which manually performs the same function as the automatic LOCA signal (i.e., ECCS Manual Initiation coincident with a Suppression Pool Water LevelLow Low will initiate the trip system).
: 8. Either trip system can actuate the LLS S/RVs by energizing the associated solenoids. Each LLS trip system is enabled and sealed in upon initial S/RV actuation from the existing reactor steam dome pressure sensors of any of the normal relief setpoint groups. The reactor steam dome pressure channels that control the opening and closing of the LLS S/RVs are arranged in either a one-out-of-one or a two-out-of-two logic depending on which LLS S/RV group is being controlled. This logic arrangement ensures that no single instrument failure can preclude the LLS S/RV function.
: 9. The LOP instrumentation causes various bus transfers and disconnects. Each Division 1 and 2 emergency bus Loss of Voltage Function is monitored by two undervoltage relays on the emergency bus and two undervoltage relays on each of the two offsite power sources. The outputs of these relays are arranged in a two-out-of-two taken three times logic configuration. Each of these relays is an inverse time delay relay. The Division 3 emergency bus Loss of Voltage Function is monitored by four undervoltage relays whose outputs are arranged in a one-out-of-two taken twice logic configuration. The output of this logic inputs to a time delay relay. Each Division 1, Division 2, and Division 3 emergency bus Degraded Voltage Function is monitored by two undervoltage relays for each emergency bus whose outputs are arranged in a two-out-of-two logic configuration. The output of this logic inputs to a time delay relay for each emergency bus.
E1-21
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-2: Example RICT Calculations RICT Tech Spec                                      LCO Condition                              Estimate(1)
(Days) 3.1.7.A        One SLC subsystem inoperable.                                                30.0 3.3.1.1.A      RPS Instrumentation - One or more Functions with one channel inoperable. 1.0 (2) 3.3.1.1.B      RPS Instrumentation - One or more Functions with two channels inoperable. 1.0 (2)
End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation - One or      30.0 3.3.4.1.A      more Functions with one required channels inoperable.
End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation - One or      30.0 3.3.4.1.B      more Functions with two channels inoperable.
ECCS Instrumentation - As required by Required Action A.1 and referenced    0.2(2) 3.3.5.1.B      in Table 3.3.5.1-1.
ECCS Instrumentation - As required by Required Action A.1 and referenced    1.4(2) 3.3.5.1.C      in Table 3.3.5.1-1.
ECCS Instrumentation - As required by Required Action A.1 and referenced    27.7 3.3.5.1.D      in Table 3.3.5.1-1.
ECCS Instrumentation - As required by Required Action A.1 and referenced    12.8 3.3.5.1.E      in Table 3.3.5.1-1.
ECCS Instrumentation - As required by Required Action A.1 and referenced    1.4(2) 3.3.5.1.F      in Table 3.3.5.1-1.
ECCS Instrumentation - As required by Required Action A.1 and referenced    1.4(2) 3.3.5.1.G      in Table 3.3.5.1-1.
RCIC System Instrumentation - As required by Required Action A.1 and        30.0 3.3.5.3.B      referenced in Table 3.3.5.3-1.
RCIC System Instrumentation - As required by Required Action A.1 and        30.0 3.3.5.3.D      referenced in Table 3.3.5.3-1.
Primary Containment and Drywell Isolation Instrumentation - One or more      30.0 3.3.6.1.A      Functions with one channel inoperable.
Primary Containment and Drywell Isolation Instrumentation - One or more      30.0 3.3.6.1.B      Functions with two channels inoperable.
Primary Containment and Drywell Isolation Instrumentation - One or more      30.0 3.3.6.1.D      required channels inoperable.
RHR Containment Spray System Instrumentation - As required by Required      30.0 3.3.6.3.B      Action A.1 and referenced in Table 3.3.6.3-1.
RHR Containment Spray System Instrumentation - As required by Required      30.0 3.3.6.3.C      Action A.1 and referenced in Table 3.3.6.3-1.
SPMU System Instrumentation - As required by Required Action A.1 and        30.0 3.3.6.4.B      referenced in Table 3.3.6.4-1.
SPMU System Instrumentation - As required by Required Action A.1 and        30.0 3.3.6.4.C      referenced in Table 3.3.6.4-1.
3.3.6.5.A      Relief and Low-Low Set Instrumentation - One trip system inoperable.        30.0 3.3.8.1.A      Loss of Power (LOP) Instrumentation - One or more channels inoperable.      30.0 3.5.1.A        One low pressure ECCS injection/spray subsystem inoperable.                  6.5 3.5.1.B        High Pressure Core Spray (HPCS) System inoperable.                          27.7 Two ECCS injection subsystems inoperable OR One ECCS injection and          0.7(2) 3.5.1.C        one ECCS spray subsystem inoperable.
3.5.1.E        One required ADS valve inoperable.                                          30.0 One required ADS valve inoperable AND One low pressure ECCS                  6.4 3.5.1.F        injection/spray subsystem inoperable.
3.5.3.A        RCIC System inoperable.                                                      30.0 3.6.1.2.C      One or more required primary containment air locks inoperable for reasons    30.0 E1-22
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-2: Example RICT Calculations RICT Tech Spec                                        LCO Condition                                Estimate(1)
(Days) other than Condition A or B.
One or more penetration flow paths with one PCIV inoperable, except due to      30.0 3.6.1.3.A      leakage not within limit.
One or more penetration flow paths with one or more primary containment        30.0 3.6.1.3.D      purge valves not within purge valve leakage limits.
3.6.1.6.A      One LLS valve inoperable.                                                      30.0 3.6.1.7.A      One RHR containment spray subsystem inoperable.                                30.0 3.6.2.3.A      One RHR suppression pool cooling subsystem inoperable.                          30.0 3.6.2.4.C      One SPMU subsystem inoperable for reasons other than Condition A or B.          30.0 3.6.5.2.C      Drywell air lock inoperable for reasons other than Condition A or B.            30.0 One or more penetration flow paths with one required drywell isolation valve    30.0 3.6.5.3.A      inoperable.
3.7.1.B        Division 1 or 2 SX subsystem inoperable.                                        11.2 3.7.6.A        Main Turbine Bypass System - Requirements of the LCO not met.                  30.0 3.8.1.A        One offsite circuit inoperable.                                                30.0 3.8.1.B        One required DG inoperable.                                                    12.8 3.8.1.C        Two offsite circuits inoperable.                                                3.2(2) 3.8.1.D        One offsite circuit inoperable AND One required DG inoperable.                  9.9 3.8.4.A        One battery charger on Division 1 or 2 inoperable.                              9.3 3.8.4.B        One battery on Division 1 or 2 inoperable.                                      3.3(2)
Division 1 or 2 DC electrical power subsystem inoperable for reasons other      2.5(2) 3.8.4.C        than Condition A or B.
3.8.7.A        Division 1 or 2 inverter inoperable.                                            30.0 One or more Division 1 or 2 AC electrical power distribution subsystems        0.0(2) 3.8.9.A        inoperable.
One or more Division 1 or 2 uninterruptible AC bus distribution subsystems      4.0(2) 3.8.9.B        inoperable.
One or more Division 1 and 2 DC electrical power distribution subsystems        3.3(2) 3.8.9.C        inoperable.
Table E1-2 Notes:
: 1. RICTs are based on the internal events, internal flood, and internal fire PRA model calculations with seismic penalties. RICTs calculated to be greater than 30 days are capped at 30 days based on NEI 06-09-A. RICTs are rounded to nearest tenth of a day.
: 2. Per NEI 06-09-A, for cases where the total CDF or LERF is greater than 1E-03/yr or 1E-04/yr, respectively, the RICT Program will not be entered for pre-planned maintenance activities.
E1-23
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-3 lists the TSTF-505 Rev 2 Table 1 Tech Specs that require additional justification along with a description of how the additional justification is provided in the LAR.
Table E1-3: TSTF-505 Rev 2 Table 1 Technical Specifications (TS) that Require Additional Justification TS Description                TSTF-505 TS      CPS TS      Additional Justification Source Range Monitor          3.3.1.2.A        3.3.1.2.A    N/A - TSTF-505 changes are excluded.
Instrumentation - One or more required SRMs inoperable in MODE 2 with intermediate range monitors (IRMs) on Range 2 or below End of Cycle Recirculation    3.3.4.1.A.1      3.3.4.1.A    Not modeled in the PRA. ATWS RPT Pump Trip (EOC-RPT)                                          will be used as a surrogate.
Instrumentation - One or more required channels inoperable                                ATWS-RPT conservatively bounds the impact of an inoperable channel by assessing the direct impact of inoperable instrumentation on the primary function served.
Under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
Low-Low-Set (LLS)            3.3.6.5.A        3.3.6.5.A    The LLS Instrumentation is not explicitly Instrumentation                                              modeled. Failure of the LLS valves themselves are used as a surrogate.
The LLS valves conservatively bound the impact of an inoperable trip system by assessing the direct impact of inoperable instrumentation on the primary function served.
Loss of Power (LOP)          3.3.8.1.A        3.3.8.1.A    LOP Instrumentation is not explicitly Instrumentation - One or more                                modeled. The function is captured and channels inoperable                                          bounded in the PRA by using the EDG start initiation circuit. Under certain circumstances, with more than one channel inoperable, a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when trip capability is not maintained.
E1-24
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-3: TSTF-505 Rev 2 Table 1 Technical Specifications (TS) that Require Additional Justification TS Description                TSTF-505 TS    CPS TS      Additional Justification Primary Containment Air Lock 3.6.1.2.C        3.6.1.2.C The Primary Containment airlocks are
- Primary containment air lock                            not explicitly modeled so a large pre-inoperable for reasons other                              existing leak failure will be used as a than Condition A or B                                      conservative surrogate for the RICT calculation. Under certain circumstances, with more than one primary containment airlock door inoperable, excessive leakage or a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when leakage exceeds limits or it there is a loss of function.
Primary Containment Isolation 3.6.1.3.E        3.6.1.3.D Not all primary containment isolation Valves (PCIVs) - One or more                              valves are modeled. Therefore, a penetration flow paths with                                surrogate of a pre-existing containment one or more containment                                    failure is chosen to conservatively purge valves not within purge                              bound the increase in risk for cases valve leakage limits.                                      where explicit modeling does not exist.
Residual Heat Removal          3.6.1.7.A      3.6.1.7.A  The Containment Spray mode of the (RHR) Containment Spray                                    RHR system is explicitly modeled in the System Condition: One RHR                                  PRA including all relevant mechanical containment spray subsystem                                components and operator actions.
inoperable.
Drywell Air lock - Drywell air 3.6.5.2.C      3.6.5.2.C The drywell air lock is not modeled in lock inoperable for reasons                                detail. Therefore, a surrogate is chosen other than Condition A or B.                              that represents a failure of containment.
Under certain circumstances, with more than one drywell airlock inoperable, excessive leakage or a loss of function may occur. Therefore, a Note is added to the Completion Time which prohibits applying a RICT when leakage exceeds limits or it there is a loss of function.
Main Turbine Bypass System    3.7.6.A        3.7.6.A    SSCs are modeled consistent with the
-                                                          TS scope and so can be directly The following limits are made                              included in the RTR tool for the RICT applicable:                                                program. The success criteria are
[a. LCO 3.2.1, "AVERAGE                                    consistent with the design basis PLANAR LINEAR HEAT GENERATION RATE                                            The TS function to limit peak pressure (APLHGR)," limits for an                                  in the main steam lines and to maintain inoperable Main Turbine                                    steam flow to the condenser for use by Bypass System, as specified                                the condensate system is the PRA in the [COLR] and ]                                        modeled function. The combined
[b. LCO 3.2.2, "MINIMUM                                    pressure control function of the turbine CRITICAL POWER RATIO                                      control valves and bypass valves while (MCPR)," limits for an                                    the main turbine is online is not inoperable Main Turbine E1-25
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions Table E1-3: TSTF-505 Rev 2 Table 1 Technical Specifications (TS) that Require Additional Justification TS Description                TSTF-505 TS    CPS TS      Additional Justification Bypass System, as specified                                modeled but this is not a mitigation in the [COLR]. ]                                          function that would affect risk.
Condition: [Requirements of the LCO not met or Main Turbine Bypass System inoperable].
E1-26
 
ENCLOSURE 1 List of Revised Required Actions to Corresponding PRA Functions
: 2. References
: 1. Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,'"
dated May 17, 2007 (ADAMS Accession No. ML071200238).
: 2. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"
Revision 0-A, dated October 12, 2012 (ADAMS Accession No. ML12286A322).
: 3. CL-PRA-003, Rev 004, "Clinton Power Station Probabilistic Risk Assessment Success Criteria Notebook", August 2018
: 4. CL-PRA-005.01, Rev 005, "Clinton Power Station Probabilistic Risk Assessment Safety Relief Valves (SRVs) and Automatic Depressurization System (ADS) System Notebook",
August 2018 E1-27
 
ENCLOSURE 2 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 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
 
ENCLOSURE 2 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2
: 1. Introduction This enclosure provides information on the technical adequacy of the Clinton Power Station (CPS) Probabilistic Risk Assessment (PRA) Internal Events model (including internal flooding) and the CPS Fire PRA model in support of the license amendment request to revise Technical Specifications to implement NEI 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines" (Reference 1).
Topical Report NEI 06-09, as clarified by the NRC final safety evaluation of this report (Reference 2), defines the technical attributes of a PRA model and its associated Configuration Risk Management Program (CRMP) tool required to implement this risk-informed application.
Meeting these requirements satisfies Regulatory Guide (RG) 1.174 (Reference 3) requirements for risk-informed plant-specific changes to a plant's licensing basis.
Exelon employs a multi-faceted approach to establishing and maintaining the technical adequacy and fidelity of PRA models for all operating Exelon nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update process and the use of self-assessments and independent peer reviews.
Section 2 of this enclosure describes requirements related to the scope of the CPS PRA models. Section 3 addresses the technical adequacy of the Internal Events PRA for this application. Section 4 similarly addresses the technical adequacy of the Fire PRA for this application. Section 5 lists references used in the development of this enclosure.
All PRA models described below have been peer reviewed, and the review and closure of F&Os from the peer review have been independently evaluated to confirm that the associated model changes did not constitute a model upgrade. Sections 3 and 4 provide the disposition of all open peer review F&Os that were associated with Supporting Requirements (SRs) assessed as "Not Met" or Capability Category (CC) I following the closure reviews, including the disposition of the open F&O relative to this application. Note that all open F&Os that represent a gap to meeting CC II are dispositioned in this application. The resolved F&Os and the basis for resolution are documented in the F&O Closure Review reports (References 11 and 12).
: 2. Requirements Related to Scope of CPS PRA Models The PRA models discussed in this enclosure have been assessed against RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2 (Reference 4) consistent with NRC RIS 2007-06 (Reference 8).
Both the CPS Internal Events PRA model (including internal flooding) and the CPS Fire PRA model are at-power models. The models are capable of quantifying Core Damage Frequency (CDF) and Large Early Release Frequency (LERF).
Note that this portion of the CPS PRA model does not incorporate the risk impacts of external events. The treatment of seismic risk and other external hazards for this application are discussed in Enclosure 4.
E2-1
 
ENCLOSURE 2 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2
: 3. Scope and Technical Adequacy of CPS Internal Events and Internal Flooding PRA Model Topical Report NEI 06-09 requires that the PRA be reviewed to the guidance of RG 1.200 Revision 2 (Reference 4) for a PRA which meets Capability Category (CC) II for the Supporting Requirements (SRs) of the American Society of Mechanical Engineers (ASME) / American Nuclear Society (ANS) PRA Standard (Reference 5). It also requires that deviations from these CCs relative to the Risk Informed Completion Time (RICT) Program be justified and documented.
The information provided in this section demonstrates that the CPS Internal Events PRA model (including internal flood) meets the expectations for PRA scope and technical adequacy as presented in RG 1.200, Revision 2.
The CPS PRA model for internal events received a formal industry peer review in October 2009 (Reference 9). The CPS Full Power Internal Events (FPIE) (including internal flooding) Peer Review was performed using the NEI 05-04 process (Reference 7), the ASME/ANS PRA Standard and Regulatory Guide 1.200, Rev. 2. The Peer Review found that 78.5% of the SRs evaluated met Capability Category II or better. There were fifty-six (56) SRs that were assessed as "Not Met" and twelve (12) SRs that were assessed as meeting only Capability Category I. Of the 68 SRs which were assessed as not meeting Capability Category II or better, seven (7) were related to Internal Flooding SRs. Several of the F&Os associated with the open SRs were related to documentation issues.
The 2009 FPIE Peer Review F&Os were addressed during several periodic PRA updates and the resolutions to the F&Os were reviewed by independent review teams in two separate F&O Closures (in December 2018 and November 2019) that included FPIE & Fire PRA F&Os (References 11 and 12). The independent review teams concluded that for the FPIE PRA, one F&O was dispositioned as "partially resolved" and one F&O was dispositioned as "open". All other F&Os representing a gap to meeting CC II for all SRs were dispositioned as "resolved".
The FPIE PRA Peer Review identified FPIE F&Os associated with SRs assessed as less than CC II. Table E2-1 summarizes those F&Os that remain "open" (including those that may be only "partially resolved") at the time of this report. The F&Os discussed in Table E2-1 represent the gaps to meeting Capability Category II for the FPIE PRA model.
As documented in Table E2-1, only two FPIE F&Os remain open. An assessment with respect to the impact on this application is also provided for each open F&O.
Based on the assessments provided in Table E2-1, it is concluded that the CPS Internal Events PRA (including internal flooding) is of adequate technical capability to support the TSTF-505 program.
E2-2
 
ENCLOSURE 2 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2 Table E2-1 CLINTON FPIE / INTERNAL FLOODING PRA PEER REVIEW OPEN FACTS AND OBSERVATIONS (POST F&O CLOSURE)
F&O  Originating                        Basis for          Possible                                                                                                    Maintenance      Impact to TSTF-505 ID    SR(s)    F&O Details            Significance      Resolution        Status  Disposition from F&O Closure Review                                              vs. Upgrade      Implementation 1-32  LE-E1    CPS-PSA-004            An inadequate      Solve with all  Partially Clinton Assessment:                                                              Clinton          This issue has minimal impact on QU-C1    Section 5.2 discusses  process to        post-initiator  Resolved  Section 5.3 and Appendix K of the Human Reliability Analysis Notebook (CL-        Assessment:      the TSTF-505 application since all the use of screening    identify          HEPs set to 1.0            PRA-004) (Reference 13) summarizes HRA Dependency Analysis methodology            Maintenance:    risk-significant HRA dependencies QU-C2 values used for HEPs    combinations of    and identify all          and results. For CDF and LERF, the FPIE model was quantified with all post-      Methodology      are captured through the current HR-H3    in order to identify    operator actions  combinations of                                                                                                              methodology and results.
initiator HEPs set to 0.1 or higher at the truncation levels of 5E-9/yr (CDF) and and tools LE-E4    cutsets with            can result in      operator action-          5E-10/yr (LERF). These truncation levels were selected because they capture      consistent with QU-A5    dependent HEPs.        significantly      related HEPs.              all risk-significant post-initiator operator actions.                            previous PRA    A review of the CDF & LERF HR-G7    However, only twelve    underestimating    Perform                                                                                                      updates.        cutsets was performed to determine of the over 100 basic  CDF and LERF.      dependency Using the HRA Calculator Dependency Module, all dependent combinations                            if any HRA dependent combinations events modeling post-                      analyses for all                                                                                                              exist without escalated dependent initiator operator                        combinations.              were reviewed for proper dependency levels and order. Once reviewed, a floor      Independent value of 1E-06 or 5E-07 may be imposed on the dependent joint HEP depending      Review Team      joint HEPs (i.e., they assume zero actions are listed in on the timing of the operator actions. The final FPIE model quantification uses  Assessment:      dependence and thus the HEPs are Table 5.2-1 as using                                                                                                                                                    unaltered). Separately, a review of screening values to                                                  the 0.1 or higher seed values for all post-initiator HEPs and the adjusted        Since no new dependent joint HEP is recovered using a post-processing recovery file.          methods were    the combinations concluded that a identify dependency.
majority of the unanalyzed Of these, six use a                                                                                                                                    applied and dependent combinations are related value of 1.0E-02 and                                                  Independent Review Team Assessment:                                              existing methods were not applied to time-phased actions (i.e., early one uses a value of                                                  A check of the CAFTA RR Database indicates that the post-initiator HEPs were                      vs. late) where no additional 1.0E-03. The                                                                                                                                            in a different set to 0.1 (or greater) prior to dependency analysis.                                              dependency need be assigned remaining five use a                                                                                                                                    context, this constitutes      between the actions because the value of 0.1. It The value of 0.1 can be acceptable depending upon what truncation level is        model            time-phased calculations already appears that all other used for the dependency analysis and whether all multiple independent HFEs                        reflect the impacts of those Hoes are quantified                                                                                                                                    maintenance.
are recovered by combination HFEs and Joint HEPs. The resolution of this                          dependencies. A few legitimate with their nominal Finding is correlated to Finding 1-34.                                                            dependent combinations were values. Use of such identified upon further review, low probability values                                                                                                                                                  however, increasing the dependent is likely to result in joint HEPs for these groups does combinations of not substantially impact the overall dependent HEPs risk results.
being omitted by truncation values.
Use of a sufficiently                                                                                                                                                    Further justification for the chosen high value for HEPs is                                                                                                                                                  truncation level used in the HRA required by SR QU-                                                                                                                                                      Dependency Analysis is required in C1 and not using a                                                                                                                                                      a future model update.
sufficiently high value would result in an                                                                                                                                                      Therefore, this open item is inadequate                                                                                                                                                              primarily a documentation assessment of                                                                                                                                                            issue.
dependent HEPs.
(This F&O originated from SR HR-G7)
E2-3
 
ENCLOSURE 2 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2 Table E2-1 CLINTON FPIE / INTERNAL FLOODING PRA PEER REVIEW OPEN FACTS AND OBSERVATIONS (POST F&O CLOSURE)
F&O  Originating                          Basis for          Possible                                                                                                    Maintenance      Impact to TSTF-505 ID    SR(s)    F&O Details              Significance      Resolution      Status Disposition from F&O Closure Review                                                  vs. Upgrade      Implementation 1-34  LE-E1    Solving the PRA          The solution      Solve the PRA  Open  Clinton Assessment:                                                                  Clinton          See discussion for F&O 1-32.
QU-C1    models with some        method used        model with            See discussion for F&O 1-32.                                                        Assessment:
HEPs at nominal can      likely under      operator action                                                                                            Maintenance:
QU-C2 result in cutsets with  predicts the risk  failure                                                                                                    Methodology HR-H3    multiple operator        values. This      probability            Independent Review Team Assessment:
and tools LE-E4    actions being            under prediction  values set to a        The CL-PRA-004 Rev. 6 document was reviewed. The final model cutsets were            consistent with QU-A5    truncated out or with    could be          high value.            re-imported into the existing HRA DAF files (for FPIE CDF only), using a copy of    previous PRA the combined            significant based                        the HRAC database with all 1.0 HEPs removed and the inhibit ADS also                updates.
HR-G7 probability of all      on the total                              removed per the analyst notes for that HFE. This process was used to operator actions much    number of                                determine if there are combinations of HFEs occurring in the final results with all below the 1E-6 or 5E-    operator actions                          HEPs set to nominal values and no combination event applied. 318 new                Independent 7 floor that the HRA    included in the                          combinations were identified (in addition to the 216 that were originally identified Review Team notebook says is        CPS model.                                and implemented), several of which had FV values above 5E-03 as calculated by        Assessment:
used. The peer                                                    the HRAC (which is not a true risk metric but a good approximation).                Maintenance -
review team                                                                                                                                            modeling error, quantified the PRA                                                For example, 1FWOPFLWCTRL-H-- and 1FWOPMANINIT-H-- appear as a                      approach will not model with post-                                                  combination together and have a dependency level of HD, confirmed in the HRA        change.
initiator HEPs set to                                              Calculator via override notes, however when this pair of HFEs appears together 0.1 and identified a                                              it is not recovered with a combination event. This combination has a an FV significant number of                                              value of 2.9E-01 as calculated by the HRAC (again, not a true risk metric but a cutsets containing                                                good approximation). This suggests it is likely risk significant when combinations of basic                                              dependencies are accounted for, and additional unanalyzed combinations may events representing                                                also be present when dependencies are accounted for.
operator action failure.
These combinations were reviewed and a                                                The review teams concern is that potentially risk significant combinations of large number of                                                    HFEs are not captured through the current approach, due to the chosen combinations                                                      truncation level for the dependency identification (5E-9 / 5E-10 for CDF/LERF) in identified in this                                                conjunction with the elevated HEP level chosen (0.1). This could under predict review were not                                                    risk results as stated in the original F&O, and is supported by the observations included in the CPS                                                noted above. It is noted that the example combination above did appear in the HRA dependency                                                    1E-9 / 5E-11 identification cutsets that were included in the dependency files, but evaluation.                                                        not used.
(This F&O originated from SR HR-G7)
E2-4
 
ENCLOSURE 2 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2 Table E2-1 CLINTON FPIE / INTERNAL FLOODING PRA PEER REVIEW OPEN FACTS AND OBSERVATIONS (POST F&O CLOSURE)
F&O    Originating                Basis for        Possible                                                                                                  Maintenance Impact to TSTF-505 ID      SR(s)    F&O Details    Significance    Resolution  Status  Disposition from F&O Closure Review                                                    vs. Upgrade Implementation 1-34                                                                  Recommendations (cont'd)                                                                Show that risk significant combinations of HFEs appearing in the final results are all captured in the dependency analysis. Some suggestions on how to accomplish this are provided below.
: 1) Include more cutsets in the dependency identification process when imported into the HRAC. The total number of cutsets generated for the dependency analysis was low (5596 / 1014) which is likely the leading cause of this issue. The final model maintained the elevated HEP values for all HFEs, suggesting model quantification time is not an issue preventing the generation of additional cutsets through lowering of the identification truncation or increasing the HEP values above 0.1. This can be accomplished by either lowering the identification truncation levels, increasing the elevated HEP values, or both. The balance between these driving factors is model specific and may require some iteration. If this approach is chosen all identified combinations can be implemented if the model allows it, however a more refined approach can be accomplished by using risk metric cutoffs to select which combinations to implement, the use of optimized seed values, or both.
: 2)    Show that the current set of combinations captures all risk significant combinations of HFEs when dependencies are accounted for through a sensitivity study on the final results. Using the final cutsets identify the unanalyzed combinations, and create additional recovery rules for them, using the conservative dependency levels automatically generated by the HRAC or refining as necessary.
Suggestion For fire this issue may also exist, as the same identification truncation levels were used, and only 21237 / 11552 cutsets were generated. After re importing the final result cutsets for Fire CDF (using an HRAC file with the 1.0s removed),
78 additional combinations were identified, of which several had FV values above 5E-03 as calculated by the HRAC. Therefore, it is suggested that the Fire dependency analysis should be revisited in a similar manner.
E2-5
 
ENCLOSURE 2 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2
: 4. Scope and Technical Adequacy of CPS Fire PRA Model The CPS Fire PRA (FPRA) Peer Review (Reference 10) was performed in April 2018 using the NEI 07-12 Fire PRA peer review process (Reference 6), the ASME/ANS PRA Standard, ASME/ANS RA-Sa-2009 (Reference 5) and Regulatory Guide 1.200, Revision 2 (Reference 4).
The purpose of this review was to establish the technical adequacy of the FPRA for the spectrum of potential risk-informed plant licensing applications for which the FPRA may be used.
The 2018 CPS FPRA Peer Review was a full-scope review of the CPS at-power FPRA against all technical elements in Part 4 of the ASME/ANS PRA Standard, including the referenced internal events Supporting Requirements (SRs). The Peer Review found that 96.9% of the SRs evaluated met Capability Category (CC) II or better. There were five (5) SRs that were assessed as "Not Met" and eight (8) SRs that were assessed as meeting only CC I. Many of the F&Os, leading to open SRs, were related to documentation issues.
The 2018 FPRA Peer Review F&Os were addressed in subsequent FPRA updates and the resolutions to the F&Os were reviewed by independent review teams in two separate F&O Closures (in December 2018 and November 2019) that included FPIE & FPRA F&Os (References 11 and 12). The independent review teams concluded that for the FPRA, all F&Os have been dispositioned as "resolved". Therefore, there are no open F&Os to discuss for this application.
Given the resolution of all F&Os related to SRs assessed with less than a CC II, it is concluded that the CPS FPRA is of adequate technical capability to support the TSTF-505 program.
E2-6
 
ENCLOSURE 2 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2
: 5. References
: 1. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"
Revision 0-A, October 12, 2012 (ADAMS Accession No. ML12286A322).
: 2. Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines'",
May 17, 2007 (ADAMS Accession No. ML071200238).
: 3. Regulatory Guide (RG) 1.174, "An Approach For Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis," Rev. 3, January 2018.
: 4. Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Rev. 2, March 2009.
: 5. 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.
: 6. NEI 07-12, "Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines,"
Rev. 1, June 2010.
: 7. NEI 05-04, "Process for Performing PRA Peer Reviews Using the ASME PRA Standard,"
Rev. 2, November 2008.
: 8. NRC Regulatory Issue Summary 2007-06, "Regulatory Guide 1.200 Implementation," March 2007.
: 9. Clinton Power Station 2009 PRA Peer Review Report, Boiling Water Reactor Owners Group (BWROG), April 2010.
: 10. Clinton Power Station Fire PRA Peer Review Report Using ASME/ANS PRA Standard Requirements, Boiling Water Reactor Owners Group (BWROG), August 2018.
: 11. Report # 032362-RPT-12, Clinton PRA Finding and Suggestion Level Fact and Observation Independent Assessment, Rev. 0, January 2019.
: 12. Report # 032434-RPT-01, Clinton Power Station Fire PRA Finding Level Fact and Observation Closure by Independent Assessment, Rev. 0, February 2020.
: 13. CL-PRA-004, Clinton Power Station Probabilistic Risk Assessment Human Reliability Analysis Notebook, Rev. 6, February 2020.
E2-7
 
ENCLOSURE 3 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Information Supporting Technical Adequacy of PRA Models Without PRA Standards Endorsed by Regulatory Guide 1.200, Revision 2 This enclosure is not applicable to the Clinton Power Station (CPS) submittal.
Exelon Generation Company, LLC is not proposing to use any PRA models in the CPS Risk-Informed Completion Time Program for which a PRA standard, endorsed by the NRC in Regulatory Guide 1.200, Revision 2 does not exist.
 
ENCLOSURE 4 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 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
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models 1    Introduction and Scope Topical Report NEI 06-09, Revision 0-A (Reference [1]), as clarified by the Nuclear Regulatory Commission (NRC) final safety evaluation (Reference [2]), requires that the License Amendment Request (LAR) provide a justification for exclusion of risk sources from the Probabilistic Risk Assessment (PRA) model based on their insignificance to the calculation of configuration risk as well as discuss conservative or bounding analyses applied to the configuration risk calculation. This enclosure addresses this requirement by discussing the overall generic methodology to identify and disposition such risk sources. This enclosure also provides the Clinton Power Station (CPS) specific results of the application of the generic methodology and the disposition of impacts on the CPS Risk Informed Completion Time (RICT)
Programs. Section 3 of this enclosure presents the plant-specific analysis of seismic risk to CPS. Section 4 of this enclosure presents the justification for excluding high wind risk to CPS.
Section 5 presents the justification for excluding External Flooding risk to CPS. Section 6 of this enclosure presents the justification for excluding other external hazards to CPS.
Topical Report NEI 06-09 does not provide a specific list of hazards to be considered in a RICT Program. However, non-mandatory Appendix 6-A in the ASME/ANS PRA Standard (Reference
[3]) provides a guide for identification of most of the possible external events for a plant site.
Additionally, NUREG-1855 (Reference [4]) provides a discussion of hazards that should be evaluated to assess uncertainties in plant PRAs and support the risk-informed decision-making process. This information was reviewed for the CPS site and augmented with a review of information on the site region and plant design to identify the set of external events to be considered. The information in the UFSAR regarding the geologic, seismologic, hydrologic, and meteorological characteristics of the site region as well as present and projected industrial activities in the vicinity of the plant were also reviewed for this purpose. No new site-specific and plant-unique external hazards were identified through this review. The list of hazards in Appendix 6-A of the PRA Standard were considered for CPS as summarized in Table E4-7.
The scope of this enclosure is consideration of the hazards in Table E4-7 for CPS. As explained in subsequent sections of this enclosure, risk contribution from seismic events is evaluated quantitatively, and the other listed external hazards are evaluated and screened as having low risk.
E4-1
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models 2    Technical Approach The guidance contained in NEI 06-09 states that all hazards that contribute significantly to incremental risk of a configuration must be quantitatively addressed in the implementation of the RICT Program. The following approach focuses on the risk implications of specific external hazards in the determination of the risk management action time (RMAT) and RICT for the Technical Specification (TS) Limiting Conditions for Operation (LCOs) selected to be part of the RICT Program.
Consistent with NUREG-1855 (Reference [4]), external hazards may be addressed by:
: 1) Screening the hazard based on a low frequency of occurrence,
: 2) Bounding the potential impact and including it in the decision-making, or
: 3) Developing a PRA model to be used in the RMAT/RICT calculation.
The overall process for addressing external hazards considers two aspects of the external hazard contribution to risk.
* The first is the contribution from the occurrence of beyond-design-basis conditions, e.g.,
winds greater than design, seismic events greater than the design-basis earthquake (DBE), etc. These beyond-design-basis conditions challenge the capability of the SSCs to maintain functionality and support safe shutdown of the plant.
* The second aspect addressed is the challenges caused by external conditions that are within the design basis, but still require some plant response to assure safe shutdown, e.g., high winds or seismic events causing loss of offsite power, etc. While the plant design basis assures that the safety-related equipment necessary to respond to these challenges are protected, the occurrence of these conditions nevertheless causes a demand on these systems that present a risk.
Hazard Screening The first step in the evaluation of an external hazard is screening based on an estimation of a bounding core damage frequency (CDF) for beyond-design-basis hazard conditions. An example of this type of screening is reliance on the NRC's 1975 Standard Review Plan (SRP)
(Reference [5]), which is acknowledged in the NRC's Individual Plant Examination of External Events (IPEEE) procedural guidance (Reference [6]) as assuring a bounding CDF of less than 1E-6/yr for each hazard. The bounding CDF estimate is often characterized by the likelihood of the site being exposed to conditions that are beyond the design-basis limits and an estimate of the bounding conditional core damage probability (CCDP) for those conditions. If the bounding CDF for the hazard can be shown to be less than 1E-6/yr, then beyond-design-basis challenges E4-2
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models from that hazard can be screened out and do not need to be addressed quantitatively in the RICT Program.
The basis for this is as follows:
* The overall calculation of the RICT is limited to an incremental core damage probability (ICDP) of 1E-5.
* The maximum time interval allowed for this RICT is 30 days.
* If the maximum CDF contribution from a hazard is <1E-6/yr, then the maximum ICDP from the hazard is <1E-7 (1E-6/yr
* 30 days/365 days/yr).
* Thus, the bounding ICDP contribution from the hazard is shown to be less than 1% of the permissible ICDP in the bounding time for the condition. Such a minimal contribution is not significant to the decision in computing a RICT.
The CPS IPEEE hazard screening analysis (Reference [7]) has been updated to reflect current CPS site conditions. The results are discussed in Section 6 and show that all the events listed in Table E4-7 can be screened, except seismic events for CPS.
Hazard Analysis - CDF There are two options in cases where the bounding CDF for the external hazard cannot be shown to be less than 1E-6/yr. The first option is to develop a PRA model that explicitly models the challenges created by the hazard and the role of the SSCs included in the RICT Program in mitigating those challenges. The second option for addressing an external hazard is to compute a bounding CDF contribution for the hazard.
Evaluate Bounding LERF Contribution The RICT Program requires addressing both core damage and large early release risk. When a comprehensive PRA does not exist, the LERF considerations can be estimated based on the relevant parts of the internal events LERF analysis. This can be done by considering the nature of the challenges induced by the hazard and relating those to the challenges considered in the internal events PRA. This can be done in a realistic manner or a conservative manner. The goal is to provide a representative or bounding conditional large early release probability (CLERP) that aligns with the bounding CDF evaluation. The incremental large early release frequency (ILERF) is then computed as follows:
ILERFHazard = ICDFHazard
* CLERPHazard E4-3
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models The approaches used for seismic LERF is described in Section 3.
Risks from Hazard Challenges Given the selection of an estimated bounding CDF/LERF, the approach considered must assure that the RICT Program calculations reflect the change in CDF/LERF caused by the out of service equipment. For CPS, as discussed later in this enclosure, the only beyond-design-basis hazard that could not be screened out is the seismic hazard, and the approach used considers that the change in risk with equipment out of service will not be higher than the estimated seismic CDF.
The above steps address the direct risks from damage to the facility from external hazards.
While the direct CDF contribution from beyond-design-basis hazard conditions can be shown to be non-significant using these steps without a full PRA, there are risks that may be addressed.
These risks are related to the fact that some external hazards can cause a plant challenge even for hazard severities that are less than the design-basis limit. For example, high winds, tornadoes, and seismic events below the design-basis levels can cause extended loss of offsite power conditions. Additionally, depending on the site, external floods can challenge the availability of normal plant heat removal mechanisms.
The approach taken in this step is to identify the plant challenges caused by the occurrence of the hazard within the design basis and evaluate whether the risks associated with these events are either already considered in the existing PRA model or are not significant to risk.
Section 3 of this enclosure summarizes the analysis for the CPS site with respect to the beyond-design-basis seismic hazard, Section 4 summarizes the analysis for the extreme winds hazard, Section 5 summarizes the analysis of the external flooding hazard, and Section 6 summarizes the analysis of the other representative external hazards for the CPS site.
E4-4
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models 3    Seismic Risk Contribution Analysis Introduction The TSTF-505 program requires accounting for seismic risk contribution in calculating extended risk informed technical specification (TS) completion times (CT, also referred to as Allowed Outage Time, AOT).
Since a seismic PRA (SPRA) was not developed for the CPS IPEEE (Individual Plant Examination for External Events) (Reference [7]), an alternative approach is taken to provide an estimate of SCDF based on the current CPS seismic hazard curve (Reference [8]), and assuming the seismic capacity of a component whose seismic failure would lead directly to core damage.
The calculation of SLERF is performed by estimating an average seismic conditional Large Early release probability (SCLERP), based on the spectrum of SCDF accident sequence types, and multiplying SCDF by the average SCLERP estimate.
CL-MISC-026, "External Hazards Assessment for Clinton Power Station," (Reference [19])
documents the seismic calculations supporting this enclosure. The following paragraphs establish the important analysis assumptions, boundary conditions and ground rules related to the calculation of SCDF and SLERF.
Input and Assumptions Hazard Curve:
The CPS seismic hazard is defined by the seismic hazard curve (SHC) provided to NRC in Reference [9], using the seismic hazard curve per Reference [8].
PGA Metric:
The ground motion metric used to define the seismic hazard in the convolution analysis is peak ground acceleration (PGA). PGA is a common ground motion metric used in seismic risk assessment analyses for nuclear power plants (Reference [10]).
E4-5
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models High Confidence of a Low Probability of Failure (HCLPF):
The assumed limiting plant seismic capacity for the Clinton plant has a high confidence of low probability of failure (HCLPF) value of 0.30g PGA, as cited in Reference [11] [GI-199, Table B.2]. This value is consistent with the CPS IPEEE review level earthquake (RLE) of 0.30g PGA as specified in Reference [7]. The uncertainty parameter for seismic capacity is represented by a composite beta factor (c) of 0.4. This is a commonly-accepted approximation and is consistent with the value used in GI-199, Table C.1, Bases for Establishing Plant-Level Fragility Curves Parameters from IPEEE Information (Reference [11]).
The Seismic Margin Assessment (SMA) was performed in response to the guidelines contained in Generic Letter 88-20 Supplement 4 and NUREG 1407. The SMA methodology, as defined in EPRI NP-6041-SL, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin" (Reference [12]), enabled development of a practical method of assessing nuclear plant seismic margins. For CPS, the Seismic Margin Earthquake (SME) assigned by the NRC is the median NUREG/CR-0098 (Reference [13]) spectrum anchored at 0.3g.
The CPS design Safe Shutdown Earthquake (SSE) is based on Regulatory Guide 1.60 (Reference [14]) response spectra with a free field ground response spectra anchored at a horizontal ground acceleration of 0.26g PGA (Section 2.5.2.6 of USAR, Reference [15]). The difference between the SSE and the SME is the "margin" being assessed in the SMA program.
It is not the intent of this program to determine the absolute largest earthquake the site could withstand, only to demonstrate with a high degree of confidence that the site could withstand an SME. When components are screened-out, it means that those items are considered to be seismically adequate for the SME evaluation. The screened-out components would have an expected high confidence of low probability of failure (HCLPF) number that is at or above the SME of 0.3g. The screening process makes it unnecessary to calculate HCLPF numbers for screened-out components.
The HCLPF capacity is intended to represent an earthquake level in which there is approximately 95% confidence of less than about a 5% failure probability. There is no further review required for screened-out components. The seismic margin capability (expressed in terms of HCLPF) for any success path is then assessed to be equal to the seismic margin capability of the weakest component in that success path. Therefore, of all the components in the success paths have a HCLPF equal to or greater than 0.3g.
Convolution to Determine SCDF:
The estimation of SCDF is performed by a mathematical convolution of the PGA-based seismic hazard curve and the Clinton PGA-based plant HCLPF from References [7] and [11]). This convolution estimation approach is a common analysis in approximating an SCDF for use in E4-6
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models risk-informed decision making (e.g., it is commonly used in RICT seismic penalty calculations; the NRC used this approach in the GI-199 risk assessment in Reference [11] in absence of a current full-scope SPRA.
SLERF:
The CPS SLERF is obtained by multiplying the calculated SCDF by an average seismic conditional large early release probability (SCLERP). The average SCLERP is estimated using information from both the Clinton FPIE PRA (Reference [16]) and fragility information from industry SPRAs.
Consideration of S-LOOP:
The analysis also assesses the incremental risk associated with seismic-induced LOOP that may occur from seismic events below the CPS seismic design basis. The analysis compares a convolution estimation of seismic-induced LOOP frequency with the random LOOP frequencies in the Clinton FPIE PRA. This analysis aspect and approach has been used in past RICT seismic penalty calculations.
Calculations The general approach to estimation of the SCDF is to use the plant level HCLPF and convolve the corresponding failure probabilities as a function of seismic hazard level with the seismic hazard curve. This is a commonly used approach to estimate SCDF when a seismic PRA is not available. This approach is the same as that used in the Vogtle pilot TSTF-505 license amendment request submittal (Reference [17]) and a previous Exelon TSTF-505 submittal for Calvert Cliffs (Reference [18]).
The key elements of the SCDF calculation are discussed below.
Seismic Hazard and Intervals The seismic hazard in units of g (PGA, peak ground acceleration) is shown in Table E4-1 (from Reference [9]). The mean fractile occurrence frequencies of Table E4-1 are used in the calculations here; use of mean values is a typical and expected PRA practice. The seismic hazard curve goes from extremely low magnitude earthquakes well below the CPS operating basis earthquake of 0.11g PGA (free field horizontal ground acceleration, Section 2.5.2.7.1 of Reference [15]) to extremely large magnitude earthquakes well beyond the CPS safe shutdown earthquake of 0.26g (free field horizontal ground acceleration, Section 2.5.2.6 of Reference
[15]).
E4-7
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models The convolution calculation of the seismic hazard curve with the Clinton PGA-based plant HCLPF fragility curve (summarized later in Table E4-2) is performed by dividing the hazard curve into seismic magnitude range intervals. In the case of the seismic hazard curve in Table E4-1, nineteen seismic hazard intervals are explicitly used in the convolution calculation and are defined by the magnitude data points (0.0005 to 0.001g; 0.001g to 0.005; and ultimately to >10g). The entire hazard curve is used in the convolution calculation for completeness but the very low magnitude data points and the very high magnitude data points are non-significant to the convolved SCDF estimate (because of very low likelihood of damage and very low likelihood of occurrence, respectively).
To facilitate calculation of the Clinton plant fragility probability at each seismic hazard interval, a representative g-level is calculated for each interval. The representative g-level for the seismic hazard intervals is calculated using a geometric mean approach (i.e., the square root of the product of the g-level values at the beginning and end of a given interval). For the last open-ended seismic interval greater than 10g, the representative g-level is estimated as 10g as opposed to a higher g-level (e.g., 11g) for modeling convenience. However, this point is immaterial given that the calculated conditional failure probability at a g-level of 10g is 1.0 and the contribution from this final interval has a negligible contribution to the overall SCDF estimate.
The seismic hazard interval annual initiating event frequency is calculated (except for the final interval) by subtracting the mean exceedance frequency associated with the g-interval (high) end point from the mean exceedance frequency associated with the g-interval beginning point.
The frequency of the last seismic hazard interval is the exceedance frequency at the beginning point of that interval. This is common practice in industry SPRAs (Reference [10]).
E4-8
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4 EPRI 2013 Seismic Hazard Data for Clinton1 AMPS(g)        MEAN          0.05        0.16          0.50        0.84        0.95 0.0005      9.59E-02      6.26E-02    7.66E-02      9.65E-02    9.93E-02    9.93E-02 0.001        8.14E-02      4.63E-02    6.17E-02      8.23E-02    9.93E-02    9.93E-02 0.005        3.38E-02      1.42E-02    2.13E-02      3.23E-02    4.56E-02    6.00E-02 0.01        1.87E-02      7.13E-03    1.07E-02      1.72E-02    2.57E-02    3.79E-02 0.015        1.24E-02      4.31E-03    6.64E-03      1.10E-02    1.72E-02    2.72E-02 0.03        5.15E-03      1.25E-03    2.07E-03      4.13E-03    7.55E-03    1.36E-02 0.05        2.20E-03      3.57E-04    6.17E-04      1.46E-03    3.47E-03    7.23E-03 0.075        9.94E-04      1.13E-04    2.04E-04      5.50E-04    1.53E-03    3.90E-03 0.1          5.40E-04      4.98E-05    9.11E-05      2.72E-04    7.77E-04    2.25E-03 0.15        2.16E-04      1.64E-05    3.14E-05      1.02E-04    3.05E-04    8.85E-04 0.3          4.09E-05      3.01E-06    6.09E-06      1.98E-05    6.54E-05    1.51E-04 0.5          1.12E-05      7.13E-07    1.55E-06      5.35E-06    1.90E-05    4.07E-05 0.75        3.71E-06      1.64E-07    4.13E-07      1.64E-06    6.26E-06    1.38E-05 1            1.61E-06      4.70E-08    1.38E-07      6.54E-07    2.72E-06    6.26E-06 1.5          4.61E-07      6.09E-09    2.42E-08      1.51E-07    7.55E-07    1.90E-06 3            4.30E-08      1.84E-10    7.89E-10      8.23E-09    6.09E-08    1.92E-07 5            5.92E-09      1.11E-10    1.44E-10      7.13E-10    6.93E-09    2.64E-08 7.5          1.02E-09      1.01E-10    1.13E-10      1.74E-10    1.08E-09    4.70E-09 10          2.65E-10      1.01E-10    1.11E-10      1.42E-10    3.19E-10    1.32E-09 1 (Reproduced from Reference [8] Table A-1a. Mean and Fractile Seismic Hazard Curves for PGA at Clinton)
E4-9
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Seismic Failure Probabilities The seismic failure probability of the CPS limiting HCLPF for each seismic hazard interval is calculated using the following fragility equations (this is for the Mean confidence level). These are the typical lognormal fragility equations used in most hazard PRAs (Reference [10]).
Fragility (i.e., failure probability) =  [ln(A/Am)/c], where is the standard lognormal distribution function A is the g level in question, Am is the median seismic capacity, and the uncertainty parameters (betas) are related as follows:
c = (u^2 + r^2)^0.5.
HCLPF and Am are related as follows:
Am = HCLPF / (exp -1.65(r + u)) or HCLPF / (exp -2.33c)
SCDF is evaluated corresponding to the HCLPF value based on the CPS IPEEE analysis, or 0.3g PGA. In the calculations for this case, the uncertainty variable c is set to a value of 0.4; this uncertainty variable value is consistent with that used by the NRC for the Clinton plant HCPLF in Reference [11] as well as it typical for use as a representative composite uncertainty (refer to Section 6.4 of Reference [20].
With all parameters specified, the hazard interval-specific CPS plant level failure probabilities are calculated as defined above. The interval-specific failure probabilities are shown in Table E4-2 for each interval (along with the hazard interval initiating event frequencies and the total convolved SCDF).
Note that the Am-to-HCLPF relationship of Am = HCLPF / (exp -2.33c), is used, as opposed to the version Am = HCLPF / (exp -1.65(r + u)), to calculate the fragility probabilities in the convolution. These two versions of the Am-to-HCLPF relationship produce almost exactly the same Am result in all fragility statistic cases, except in the less common cases where the r and u are significantly different in value.
E4-10
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Seismic Core Damage Frequency The SCDF for each hazard interval is then the product of the hazard interval initiating event frequency (/yr) and the plant level HCLPF fragility failure probability for that same hazard interval. The results per hazard interval are then straight summed to produce the overall total SCDF across the entire hazard curve. The SCDF convolution calculation is summarized in Table E4-2 and shows the total estimated SCDF is 6.4E-6/yr. Table E4-2 provides the following information:
* CPS limiting seismic plant HCLPF inputs
* Seismic hazard intervals and their associated initiating event frequencies (Mean) and representative magnitudes
* Plant level HCLPF fragility failure probabilities (Mean) per hazard interval
* Convolved SCDF per interval and total SCDF.
Seismic Large Early Release Frequency For use of a seismic penalty estimate in RICT calculations, it can be unacceptably conservative to simply assume that Delta SCDF = Delta SLERF. As such, a less conservative approach is pursued here for the CPS SLERF estimation. The CPS SLERF is determined by multiplying the estimated SCDF shown in Table E4-2 (6.4E-06/yr) by an average seismic conditional large early release probability (SCLERP). An estimate of the average SCLERP is calculated using 1) an estimation of the breakdown of SCDF by accident sequence type and 2) PRA accident sequence progression information from the quantification results of the current CPS FPIE PRA model of record (Reference [16]) adjusted to reflect the influence of seismic-induced failures.
E4-11
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-2 Convolution Calculation Summary of Clinton Seismic CDF CPS Limiting Plant HCLPF    CPS Seismic Hazard Curve                            Convolution Calculation (IPEEE SMA, Reference [7])                                      (CPS limiting plant HCLPF fragility with Seismic Hazard)
Hazard Mean            Hazard Interval Am            Peak Ground                                        Hazard Interval      Interval      Convolved HCLPF                                  Exceedance        Representative (g,    c    Acceleration                                            Fragility      Occurrence      Frequency (g, PGA)                                  Frequency          Magnitude PGA)                (g)                                                (Mean)        Frequency          (/yr)
(/yr)        (geo. mean, g PGA)
(/yr) 0.3    0.76    0.40    0.0005      9.59E-02            7.07E-04            1.55E-68          1.45E-02        2.25E-70 0.001      8.14E-02            2.24E-03            1.95E-48          4.76E-02        9.27E-50 0.005      3.38E-02            7.07E-03            6.42E-32          1.51E-02        9.69E-34 0.01      1.87E-02            1.22E-02            2.68E-25          6.30E-03        1.69E-27 0.015      1.24E-02            2.12E-02            1.73E-19          7.25E-03        1.26E-21 0.03      5.15E-03            3.87E-02            4.74E-14          2.95E-03        1.40E-16 0.05      2.20E-03            6.12E-02            1.46E-10          1.21E-03        1.77E-13 0.075      9.94E-04            8.66E-02            2.72E-08          4.54E-04        1.24E-11 0.1      5.40E-04            1.22E-01            2.44E-06          3.24E-04        7.91E-10 0.15      2.16E-04            2.12E-01            6.96E-04          1.75E-04        1.22E-07 0.3      4.09E-05            3.87E-01            4.54E-02          2.97E-05        1.35E-06 0.5      1.12E-05            6.12E-01            2.92E-01          7.49E-06        2.19E-06 0.75      3.71E-06            8.66E-01            6.26E-01          2.10E-06        1.31E-06 1        1.61E-06            1.22E+00            8.82E-01          1.15E-06        1.01E-06 1.5      4.61E-07            2.12E+00            9.95E-01          4.18E-07        4.16E-07 3        4.30E-08            3.87E+00            1.00E+00          3.71E-08        3.71E-08 5        5.92E-09            6.12E+00            1.00E+00          4.90E-09        4.90E-09 7.5      1.02E-09            8.66E+00            1.00E+00          7.55E-10        7.55E-10 10        2.65E-10            1.00E+01            1.00E+00          2.65E-10        2.65E-10 Total Convolved SCDF Across Hazard Curve (1/yr):                  6.4E-06 E4-12
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models This SLERF methodology is discussed below according to the following topics:
* Spectrum of seismic-induced core damage accident sequence types
* CLERP as a function of seismic core damage accident sequence type
* Application of SLERF in RICT Calculations Spectrum of Seismic-Induced Core Damage Accident Sequence Types The estimation of an average SCLERP requires as an input the assessment of the contribution of different accident sequence types to seismic core damage frequency (SCDF). The contribution of various accident sequence types (or accident classes) to core damage frequency at a given plant is not necessarily the same between FPIE PRA and other hazard (e.g., seismic) PRAs. Given that Clinton does not have a detailed plant-specific seismic PRA to assist in estimating the spectrum of SPRA accident sequence types and SCLERP, the range of fragility information from other recently submitted SPRAs is reviewed to assist in the calculation of SCLERP.
Based on past and current SPRAs (e.g., References [21] to [34]), the spectrum of SCDF sequence types are as follows:
* Seismic-LOOP with early loss of injection: These are seismic-induced loss of offsite power scenarios with RPV coolant injection failure at t=0.
* Seismic-LOOP with loss of containment cooling: These are seismic-induced loss of offsite power scenarios with RPV coolant makeup initially successful but containment cooling (e.g., RHR) is not successful. Adequate core cooling is subsequently failed due to containment overpressurization and failure and the subsequent effect on RPV injection equipment.
* Seismic-LOOP with Seismic-LOCA and early loss of injection: These are scenarios with a seismic-induced LOOP and seismic-induced LOCA (small, medium or large) and RPV coolant injection failure at t=0.
* Seismic-LOOP with Seismic-LOCA and loss of containment cooling: These are scenarios with a seismic-induced LOOP and seismic-induced LOCA (small, medium or large) with RPV coolant makeup initially successful but containment cooling (e.g., RHR) is not successful. Adequate core cooling is subsequently failed due to containment overpressurization and failure and the subsequent effect on RPV injection equipment.
* Seismic-ATWS Unmitigated: These are seismic-induced failure to scram scenarios with failure of reactivity control (e.g., failure of standby liquid control). These accidents proceed with high reactor power discharge into containment resulting in dynamic loading and failure of the containment structure. Adequate core cooling is subsequently failed due to containment failure and the subsequent effect on RPV injection equipment. [Note:
E4-13
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Given the very low probability of random failure to scram, seismic-induced events with random failure to scram are encompassed by this accident class category.]
* Direct to Core Damage: These are scenarios with significant seismic-induced failures that are modeled directly as core damage. Such scenarios include key structural failures (e.g., RPV support failure, reactor building and control building structural failure) and ISLOCA scenarios. This category of accidents can be further subdivided as: 1) seismic-induced failures that are often modeled in SPRAs directly as SCDF and SLERF (e.g.,
containment structural failures, containment bypass scenarios); and 2) seismic-induced failures that are often modeled in SPRAs directly as SCDF but not necessarily direct to SLERF (e.g., failure of a control building would effectively create a loss of injection at t=0 scenario but such a scenario would not lead directly to LERF).
* Seismic-Transients (no LOOP): These seismic-initiated accident sequences have offsite power available (and thus potential use of balance of plant (BOP) equipment, e.g.,
Feedwater), and RPV coolant injection failure at t=0. This category can include loss of injection scenarios or loss of containment heat removal scenarios. Given the low seismic-capacity of offsite power equipment in comparison to the much higher seismic capacities of the NSSS system and the scram system, the contribution from seismic-transients with seismic-induced failure to scram or a seismic-induced LOCA is negligible.
The above accident sequence categories cover the key critical safety functions (reactivity control, core cooling, RPV and containment integrity) and are sufficient to describe the spectrum of SCDF accident sequences. Determining the estimated fractional contribution to SCDF from each of the accident sequence types is approached as follows:
* Similar to a seismic initiating event tree structure, consider the more severe sequence types first and the remainder proceed to lesser severe accident states
* Information from industry SPRAs is used to inform the selection of key SSC fragilities Figure E4-1 shows in graphical format the approach to estimating the fraction contribution to SCDF by accident sequence type. Figure E4-1 is a seismic-initiating event tree that begins with severe seismic-induced failures that would typically be modeled as directly to SCDF and SLERF (i.e., the down branch of the "No Direct LERF" node) or directly to SCDF but not directly to SLERF (i.e., the down branch of the "No Direct CD" node). The next node, LOOP, models seismic-induced loss of offsite power. The next node, SCRAM, models seismic-induced failure to scram and the final node, LOCA, models seismic-induced LOCA.
A convolution calculation of the Table E4-1 seismic hazard curve and the fragility Am value shown at the branch points is performed for each of the key sequences shown in the SIET of Figure E4-1:
E4-14
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Figure E4-1 Contribution of SCDF By Accident Sequence Type FRACTION of NO            NO SEISMIC                                                                    ACCIDENT  FRACTION of        SCDF DIRECT        DIRECT        LOOP      SCRAM            LOCA  #
EVENT                                                                        TYPE        SCDF          (Nominal LERF            CD Fraction)
S-IE  S-DCDLRF        S-DCD        S-LOOP    S-ATWS        S-LOCA 1  S-TRANS  non-significant non-significant 2    S-LOOP S-LOOP or          0.40 S-LOOP-LOCA Am (g) = 0                        3 S-LOOP-LOCA Am (g) = 1.25        4    S-ATWS        0.10            0.10 Am (g) = 1.00                                5    S-DCD        0.37            0.35 Am (g) = 1.50                                              6  S-DCDLRF        0.16            0.15
* S-DCDLRF Node: Down at this node produces an accident sequence type that is typically and reasonably modeled in an SPRA as directly to core damage and directly to LERF (thus, CLERP=1.0). The SSC fragility for this node are the containment structure or other significant failures that would result in containment bypass (e.g., RPV supports for a BWR). A median capacity of 1.5g PGA is selected to model this node based on a review of industry SSC fragility information. This capacity is reasonably judged to be on the low end of the capacity range given the range of industry fragilities as well as the newer vintage and design (SSE=0.26g) of the Clinton plant (Reference [19]). The uncertainty parameter for seismic capacity is represented by a composite beta factor (c) of 0.4; which is a commonly accepted approximation.
* S-DCD Node: Down at this node produces an accident sequence type that is typically and reasonably modeled in an SPRA as directly to core damage but not typically directly to LERF. The SSC fragility for this node are structures such as a diesel generator building or control building; such failures would effectively result in loss of all RPV injection at t=0 and no recovery potential and would result directly in core damage but the condition probability of producing LERF would be less than 1.0. A median capacity of 1g E4-15
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models PGA is selected to model this node based on a review of industry SSC fragility information. This capacity is reasonably judged to be on the low end of the capacity range given the range of industry fragilities as well as the newer vintage and design (SSE=0.26g) of the Clinton plant (Reference [19]). The convolution calculation for accident sequence type #5 in Figure E4-1 includes the success probability for the up branch of the previous node.
* S-LOOP Node: If the S-DCDLRF and S-DCD failures did not occur this node then considers loss of offsite power. For the purposes of calculating the seismic penalty, the seismic capacity of the offsite power function is conservatively assigned an Am=0g capacity, i.e., probability of failure of 1.0 for an earthquake of any size. This has the effect of no SCDF contribution from the S-Transients (i.e., accidents with offsite power intact) which is an accident type with a lower CLERP value than the other accident types in Figure E4-1.
* S-ATWS Node: This node models the probability of a seismic-induced scram due to seismic-induced failure of reactor internals causing mechanical failure of control rods to insert. A median capacity of 1.25g PGA is selected to model this node based on a review of industry SSC fragility information. This capacity is reasonably judged to be on the low end of the capacity range given the range of industry fragilities as well as the newer vintage and design (SSE=0.26g) of the Clinton plant (Reference [19]). The convolution calculation for accident sequence type #4 in Figure E4-1 includes the success probability for the up branch of the previous nodes.
* Sequence Types #2 and #3: If the S-DCDLRF, S-DCD and S-ATWS failures did not occur then the potential sequences proceed to a seismic-induced LOOP state with potential additional failures such as a LOCA. The fractional SCDF risk contribution for these remaining sequence types is calculated as 1.0 minus the sum of the fractional contributions from the other sequence types.
The above approach to estimating the fractional contribution of accidents sequence types to SCDF is reasonable and conservative for the purpose of seismic penalty RICT calculation. The two main conservatisms are as follows:
* The lower range of seismic capacity is assigned to the nodal branch points in Figure E4-1 to force higher fractional contributions to those accident sequence types with higher CLERPs.
* Direct summation (and subtraction to determine the S-LOOP contribution), as opposed to Boolean algebra, is used to determine the accident sequence contributions to SCDF. If a Boolean summation approach were used the risk contribution from S-DCD and E4-16
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models S-DCDLRF in a full SPRA it would be much less than the combined contribution of 50%
shown in Figure E4-1 and used in the risk calculation.
CLERP as a Function of SCDF Accident Sequence Type The next step in the estimation of an average seismic CLERP is to estimate the CLERP for each SCDF accident sequence type. A given accident sequence type may not result in a core damage event until well after the PRA "Early" release time frame (defined in the CPS FPIE PRA as  4 hours from the time of the cue for a General Emergency declaration; per Section 5.5 of Reference [35].
Conversely, some accident sequence types would, by PRA convention, be modeled directly as a LERF, such as a station blackout scenario with failure to manually isolate containment isolation valves that are initially open and do not automatically isolate. Seismic CLERP as a function of SCDF accident sequence type is summarized in Table E4-3 and discussed below.
* Seismic-induced LOOP with early loss of coolant injection is assigned SCLERP of 1.1E-01. This SCLERP estimate is based on results from the current Clinton at-power internal events PRA for LOOP scenarios with no injection at t=0 and no AC recovery and no coolant injection recovery. This CLERP result is applicable to a seismic-induced accident sequence with loss of all injection at t=0 and no recovery given that the probability of the LERF release is a function of the accident progression, energetic severe accident phenomena (e.g., hydrogen explosion, steam explosion) and primary containment isolation. The Clinton Mark III containment design does not have a primary containment steel shell with an air gap (such as in Mark I containments) and thus the likelihood of a "High" magnitude release for an unmitigated core damage accident is lower in comparison to a Mark I containment design. In addition, refer to later discussion on the negligible likelihood of seismic-induced failure of containment or containment isolation.
* Seismic-induced failures leading directly to core damage (but not directly to LERF) are assigned an SCLERP based on conservatively assuming all these scenarios are loss of RPV injection at t=0 with no recovery. As such, the SCLERP of 1.1E-01 used above for the S-LOOP accidents would be applicable to the S-DCD accidents as well.
* Seismic-induced failures leading directly to core damage and LERF are, by definition, assigned an SCLERP of 1.0. There scenarios include seismic-induced failures such as containment structural failures, RPV support failures, ISLOCA and containment isolation failures.
* Unmitigated ATWS scenarios are assigned an SCLERP of 2.1E-01 based on the results of the CL2017B Level 2 PRA results. The CL2017B ATWS Level 2 sequences were reviewed in detail to confirm the applicability of this ATWS CLERP. Based on review of the CL2017B Level 2 PRA ATWS cutsets, the CL2017B FPIE-based ATWS CLERP is E4-17
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models non-significantly influenced by systemic failures whose probabilities would potentially be influenced by seismic events.
* Seismic-induced LOOP with initial coolant injection but loss of containment heat removal, or seismic-induced LOOP coincident with seismic-induced LOCA: As summarized previously and in Table E4-3, these sub-categories of S-LOOP are conservatively included in the seismic penalty averaged SCLERP estimate in the S-LOOP with early loss of RPV injection contribution discussed above. These scenarios are small risk contributors to the seismic-induces LOOP fraction as well as these sub-categories have lower CLERPs than the 1.1E-01 used for the S-LOOP with early loss of RPV injection category.
* Seismic-Transients (offsite power available): As summarized previously and in Table E4-3, seismic-induced transients are very low risk contributors to seismic-induced risk and also have a lower CLERP than the other accident sequence types used in the average SCLERP calculation shown in Table E4-3.
The sequence weighted average SCLERP over the SCDF accident sequence contributions and assigned SCLERPs is estimated as 0.25.
E4-18
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-3 SPECTRUM OF SCDF ACCIDENT SEQUENCES AND ASSOCIATED SCLERP L1 SPRA Accident
                            %SCDF(1) SCLERP(2)                      Comment Sequence Type S-LOOP: Seismic-            0.40      0.11      Based on CLERP results for LOOP with no induced loss of offsite                          injection at t=0 accidents with no AC recovery power.                                          (i.e., Class IBE) and no coolant injection recovery in the Clinton FPIE Level 2 PRA.
Clinton containment does not have a steel shell liner with an air gap (such as in Mark I containments) and thus likelihood of a "High" magnitude release for an unmitigated core damage accident is lower in comparison to a Mark I containment design.
This accident category for the purposes of the seismic penalty RICT calculation includes, in addition to the SBO with no AC recovery, those S-LOOP sequences with EDG in operation but subsequent injection failure or S-LOOP scenarios also including seismic-induced LOCAs. The SCLERP for early SBO with no recovery is higher in probability than these sub-sets of the S-LOOP accident type.
S-DCD: Scenarios            0.35      0.11      These direct to core damage (but not direct to core damage                            SLERF) accident scenarios are scenarios but not direct LERF                              that characterized as loss of all injection at (e.g., CB structural                            t=0, no injection recovery and containment failures)                                        not seismically failed. The SCLERP of 0.11 used above for the S-LOOP accidents would be applicable to the S-DCD accidents as well.
S-DCDLRF: Scenarios        0.15      1          By definition, the accidents scenarios defined direct to LERF (e.g.,                            by seismic-induced failures that would be containment structural                          modeled as direct LERF receive an SCLERP failures)                                        probability of 1.0.
E4-19
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-3 SPECTRUM OF SCDF ACCIDENT SEQUENCES AND ASSOCIATED SCLERP L1 SPRA Accident
                        %SCDF(1) SCLERP(2)                      Comment Sequence Type S-ATWS: S-ATWS          0.10      0.21      Based on review of the Clinton 2017B Level unmitigated                                  2 PRA, the majority of unmitigated ATWS scenarios would be either Moderate/Early or Low/Early given the likely containment failure location is above the suppression pool water line and would be scrubbed. The 0.21 SCLERP is for those ATWS sequences that would result in H/E without credit for injection systems (refer to Table 3.4-6A of CL-PRA-013, Rev. 6 (Reference [36]). The CL2017B FPIE-based CLERP of 0.21 is applicable for use as-is in the CPS RICT seismic penalty calculation of ATWS SLERF contribution given the ATWS LERF cutsets are primarily phenomenological and containment failure location probabilities.
S-LOOP with long term    Note (3)  Note (3)  Declaration of a general emergency would be loss of containment                          in accordance with CPS Emergency Action cooling                                      Levels. However, the Clinton PRA includes a 5% probability that the General Emergency declaration is delayed and thus can result in an "Early" release for these sequences (refer to Appendix I of the Clinton Level 2 PRA notebook (Reference [35]). Using a 5E-02 SCLERP value would be conservative because it would not account for the containment failure location in reducing release magnitude (i.e., if failure occurs above the suppression pool water line the release would be scrubbed and not a "High" magnitude release). The 0.05 SCLERP is lower than the S-LOOP SCLERP above of 0.11 and this sub-category of S-LOOP is conservatively included in the seismic penalty averaged SCLERP estimate in the S-LOOP contribution above.
E4-20
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-3 SPECTRUM OF SCDF ACCIDENT SEQUENCES AND ASSOCIATED SCLERP L1 SPRA Accident
                            %SCDF(1) SCLERP(2)                          Comment Sequence Type S-LOOP with S-LOCA          Note (3)    Note (3)  SCLERP would be similar to CPS FPIE with early loss of                                  LOOP early loss of injection case above injection                                            except the probabilities of containment failure due to certain energetic phenomena (e.g.,
direct containment heating; high pressure blowdown overwhelming vapor suppression) are much lower likelihood (or even precluded) given the LOCA condition. As such, this sub-category of S-LOOP including a seismic-induced LOCA is conservatively included in the seismic penalty averaged SCLERP estimate in the S-LOOP contribution above.
S-LOOP with S-LOCA          Note (3)    Note (3)  Same basis discussed above for S-LOOP and long-term loss of                                with loss of containment cooling. This sub-containment cooling                                  category of S-LOOP is conservatively included in the seismic penalty averaged SCLERP estimate in the S-LOOP contribution above.
S-Transients (no LOOP)      0.00        n/a        S-Transients (no seismic-induced LOOP) are reasonably assumed to be non-significant contributors to SCDF and SLERF. This is typical of past SPRAs and due in large part to the comparatively very low seismic capacity of offsite power equipment (primarily ceramic insulators).
Sequence-Weighted Average                0.25      Sum of (%SCDF x SCLERP) over all SCLERP:                                              sequence types Notes to Table E4-3:
(1)    Range of key SSC fragilities in industry SPRAs used to inform SCDF breakdown by accident type.
(2)    These are CL 2017B FPIE PRA-based CLERP estimates that are reviewed and adjusted, if necessary, to reflect seismic considerations (e.g., no credit for recovery of offsite power or injection), yielding the "SCLERP" label.
E4-21
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models (3)  These SCDF accident sequence types are variants of the S-LOOP category and are addressed in this average SCLERP estimate by conservatively incorporating them into the S-LOOP category, as summarized in the Comment column.
E4-22
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models In addition to the average SCLERP estimation discussed previously, the following discussions regarding random and seismic-induced failure of containment isolation failure are provided to support the reasonableness of the average SCLERP estimation (e.g., there are no normally-open AC-powered MOV PCIVs that would lead directly to an unscrubbed release and a LERF end state):
* Containment Isolation Random Failure: Random failure of primary containment isolation is already included in the average SCLERP estimation discussed previously.
* Containment Isolation Fragility: Seismic-induced failure of containment isolation is very low likelihood and encompassed by the SCLERPs used in Table E4-3. The containment isolation valves (PCIVs) modeled in the Clinton L2 PRA containment isolation fault tree are summarized in Table E4-4. Note that the containment isolation fault tree also includes contributions for pre-existing containment leakage and various containment hatches not properly closed (the probabilities of these potential pre-existing contributors are not influenced by the seismic event). As can be seen from Table E4-4, the containment isolation valves of interest to the LERF risk metric are air-operated valves (AOVs), most normally-closed at-power, that fail-safe closed on loss of pneumatic or electric power (e.g., seismic-induced LOOP). Successful primary containment isolation in preventing a LERF release for seismic-induced accidents is not dependent upon pneumatic supply, electric power, or containment isolation signals (i.e., ~99% of SCDF involves seismic-induced LOOP and the PCIVs fail-safe closed under such conditions).
The PCIVs have very high seismic capacities such that seismic loading will have a negligible likelihood of failing the PCIVs in the open position. These PCIVs are AOVs that fail-safe closed via internal spring force inside the AOV operator. Once closed, these valves do not need to open again during or after the seismic event. Therefore, they do not meet the definition of an "active" valve per the air operated valve equipment class (per the EPRI SQUG Generic Implementation Procedure, GIP, and EPRI NP-7149 Seismic Adequacy of Equipment Classes). The spring will successfully cause the PCIVs to shut at accelerations much greater than those associated with the functional failure capacity used to determine the fragility of active valves. As such, these PCIVs are essentially inactive valves, which are inherently rugged as there is not a credible seismic failure mechanism that would prevent the valves from failing shut as desired. In addition, both in-series AOV PCIVs in a penetration line would have to seismically fail to fail-safe closed to result in an open release pathway.
E4-23
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Some primary containment penetrations use motor operated valves (MOV) for containment isolation which would require electric power for closure and for an isolation signal. However, such PCIV MOVs are not significant to LERF for one or more of the following reasons:
MOV in closed position during at-power operation and at the time of the seismic event (e.g., main steam line drains)
Very small line (e.g., 1" diameter instrument gas line)
AOV or check valve PCIV in-series with the MOV Penetration is a closed-loop system or otherwise scrubbed (e.g., to spent fuel pool) that would not represent a LERF (i.e., High magnitude release).
Summary of Seismic LERF Calculation Based on the information in Table E4-3, an estimate of average CLERP for use in seismic penalty RICT calculations is 0.25. Therefore, a seismic penalty of SLERF is calculated as:
SLERF = 6.4E-6/yr (SCDF from Table E4-2) x 0.25 (average SCLERP from Table E4-3)
                    = 1.6E-6/yr For BWR plants with a Mark I or II containment, if a RICT is being entered during a period when the containment is de-inerted, a different SLERF penalty (i.e., consistent with a CLERP = 1.0) may be applied to address the increased potential for hydrogen deflagration events given oxygen in the containment. This approach is not applicable for Clinton which has a different containment design and is not inerted with nitrogen. As the Clinton Mark III containment is not inerted during normal at-power operation (or at any time), hydrogen igniters in Mark III containments are used to burn hydrogen to prevent explosive levels of hydrogen developing post-core damage. During postulated post-core damage scenarios without AC power the probability of hydrogen explosion induced containment failure increases significantly. This aspect is already included in the average SCLERP calculation (i.e., the SCLERP calculations include the increased probabilities of hydrogen explosion induced containment failure given no AC power to the igniters).
Application of SLERF in RICT Calculations The SLERF estimate documented above is conservatively used in the RICT process. Conservatism in the RICT process derives from the proposed approach to apply the total estimated annual seismic LERF as a delta SLERF in each RICT calculation, regardless of the duration of the completion time.
The total estimated annual seismic CDF and LERF will be applied starting at time zero for each RICT calculation.
E4-24
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-4
 
==SUMMARY==
OF CLINTON LEVEL 2 PRA PRIMARY CONTAINMENT ISOLATION Penetration Pathway Penetration(1)                  Valve ID    Valve Type            SPRA Comment Normal Status Main Steam      Normally open  1B21-F022A    AOV            Loss of pneumatic force to Line A          at-power                                      AOVs upon seismic-LOOP 1B21-F028A    AOV (PC isolation signal not necessary). AOVs fail-safe closed and not required to re-open. As such, fragility would be assessed as "rugged".
Main Steam      Normally open  1B21-F022B    AOV            Loss of pneumatic force to Line B          at-power                                      AOVs upon seismic-LOOP 1B21-F028B    AOV (PCIV signal not necessary).
AOVs fail-safe closed and not required to re-open. As such, fragility would be assessed as "rugged".
Main Steam      Normally open  1B21-F022C    AOV            Loss of pneumatic force to Line C          at-power                                      AOVs upon seismic-LOOP 1B21-F028C    AOV (PCIV signal not necessary).
AOVs fail-safe closed and not required to re-open. As such, fragility would be assessed as "rugged".
Main Steam      Normally open  1B21-F022C    AOV            Loss of pneumatic force to Line D          at-power                                      AOVs upon seismic-LOOP 1B21-F028C    AOV (PCIV signal not necessary).
AOVs fail-safe closed and not required to re-open. As such, fragility would be assessed as "rugged".
E4-25
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-4
 
==SUMMARY==
OF CLINTON LEVEL 2 PRA PRIMARY CONTAINMENT ISOLATION Penetration Pathway Penetration(1)                    Valve ID    Valve Type          SPRA Comment Normal Status Drywell Purge    Normally        1VQ001A      AOV            Loss of pneumatic force to Supply          closed at-                                    AOVs upon seismic-LOOP power            1VQ001B      AOV (PCIV signal not necessary).
AOVs fail-safe closed and At-power PRA                                  not required to re-open. As uses 1E-4                                    such, fragility would be probability that                              assessed as "rugged".
operator has DW purge open at time of event Drywell Purge    Normally        1VQ002        AOV            Loss of pneumatic force to Exhaust (Vent)  closed at-                                    AOVs upon seismic-LOOP power            1VQ005        AOV (PCIV signal not necessary).
AOVs fail-safe closed and At-power PRA                                  not required to re-open. As uses 1E-4                                    such, fragility would be probability that                              assessed as "rugged".
operator has DW purge open at time of event Containment      Normally        1VR001A      AOV            Loss of pneumatic force to Purge 36" Supply closed at-                                    AOVs upon seismic-LOOP power            1VR001B      AOV (PCIV signal not necessary).
AOVs fail-safe closed and At-power PRA                                  not required to re-open. As uses 1E-4                                    such, fragility would be probability that                              assessed as "rugged".
operator has Containment purge open at time of event E4-26
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-4
 
==SUMMARY==
OF CLINTON LEVEL 2 PRA PRIMARY CONTAINMENT ISOLATION Penetration Pathway Penetration(1)                    Valve ID    Valve Type          SPRA Comment Normal Status Containment      Normally        1VQ004A      AOV            Loss of pneumatic force to Purge 36"        closed at-                                    AOVs upon seismic-LOOP Exhaust (Vent)  power            1VQ004B      AOV (PCIV signal not necessary).
AOVs fail-safe closed and At-power PRA                                  not required to re-open. As uses 1E-4                                    such, fragility would be probability that                              assessed as "rugged".
operator has DW purge open at time of event Containment      Normally open 1VR006A          AOV            Loss of pneumatic force to Ventilation 12"  at power                                      AOVs upon seismic-LOOP Supply                            1VR006B      AOV (PCIV signal not necessary).
AOVs fail-safe closed and not required to re-open. As such, fragility would be assessed as "rugged".
Containment      Normally open    1VQ007A      AOV            Loss of pneumatic force to Ventilation 12"  at power                                      AOVs upon seismic-LOOP Exhaust                          1VQ007B      AOV (PCIV signal not necessary).
AOVs fail-safe closed and not required to re-open. As such, fragility would be assessed as "rugged".
Containment      Normally open    1F005A        AOV            Loss of pneumatic force to Instrument Air  at power                                      AOVs upon seismic-LOOP Supply                            1F006A        AOV (PCIV signal not necessary).
AOVs fail-safe closed and not required to re-open. As such, fragility would be assessed as "rugged".
E4-27
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-4
 
==SUMMARY==
OF CLINTON LEVEL 2 PRA PRIMARY CONTAINMENT ISOLATION Penetration Pathway Penetration(1)                  Valve ID    Valve Type            SPRA Comment Normal Status Drywell Floor  Normally open  1RF019        AOV            Loss of pneumatic force to Drain          at power                                      AOVs upon seismic-LOOP 1RF020        AOV (PCIV signal not necessary).
AOVs fail-safe closed and not required to re-open. As such, fragility would be assessed as "rugged".
Some side lines off the DW floor drain are equipped with check valves that are also seismically rugged.
Containment    Normally open  1RF021        AOV            Loss of pneumatic force to Floor Drain    at power                                      AOVs upon seismic-LOOP 1RF022        AOV (PCIV signal not necessary).
AOVs fail-safe closed and not required to re-open. As such, fragility would be assessed as "rugged".
Drywell        Normally open  1RE019        AOV            Loss of pneumatic force to Equipment Drain at power                                      AOVs upon seismic-LOOP 1RE020        AOV (PCIV signal not necessary).
AOVs fail-safe closed and not required to re-open. As such, fragility would be assessed as "rugged".
Some side lines off the DW equipment drain are equipped with check valves that are also seismically rugged.
E4-28
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-4
 
==SUMMARY==
OF CLINTON LEVEL 2 PRA PRIMARY CONTAINMENT ISOLATION Penetration Pathway Penetration(1)                    Valve ID    Valve Type          SPRA Comment Normal Status Containment      Normally open    1RE021        AOV          Loss of pneumatic force to Equipment Drain  at power                                      AOVs upon seismic-LOOP 1RE022        AOV (PCIV signal not necessary).
AOVs fail-safe closed and not required to re-open. As such, fragility would be assessed as "rugged".
RHR Fuel Pool    Normally          1FC036        MOV          The MOVs would not have Cooling Assist  closed at-                                    power during a seismic-(Supply)        power            1FC037        MOV induced SBO; however, At-power PRA                                  these valves are normally uses 1E-2                                    closed at power and also probability that                              would be a scrubbed operator has                                  release through the fuel pool this line open at                            (refer to CPS P&ID M05-time of event                                1037 Sheet 0001, Rev. W, Fuel Pool Cooling and Clean-up)
Upper Pool to    Normally open    1FC007        MOV          The MOVs would not have Surge Tank      at power.                                    power during a seismic-1FC008        MOV induced SBO. The PRA models an operator action to locally manually close the outboard MOV; however, this pathway does not result directly in a LERF release (refer to CPS P&ID M05-1037 Sheet 0001, Rev. W (Reference [39]), Fuel Pool Cooling and Clean-up, and Sheet 002 Rev. AE (Reference [40]).
E4-29
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-4
 
==SUMMARY==
OF CLINTON LEVEL 2 PRA PRIMARY CONTAINMENT ISOLATION Penetration Pathway Penetration(1)                  Valve ID    Valve Type            SPRA Comment Normal Status Pre-Existing    This is a PRA- Various valves Various        The 7.7E-3 probability of a Containment    modeled latent and seals                      pre-existing significant Failure(2)      condition for                                leakage from the DW or pre-existing                                  containment is non-significant                                  significant compared to the containment                                  CLERPs used in the leakage from                                  analysis and non-either the                                    significantly impacts the drywell or the                                calculated average containment                                  SCLERP. A seismic event and each is                                  does not impact the latent modelled with a                              probability of this postulated 7.7E-3                                        leakage.
probability.
E4-30
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-4
 
==SUMMARY==
OF CLINTON LEVEL 2 PRA PRIMARY CONTAINMENT ISOLATION Penetration Pathway Penetration(1)                            Valve ID    Valve Type            SPRA Comment Normal Status Drywell Hatches      The CPS PRA Various                Various    Incorrect sealing (or closure) of not Properly          also models        Hatches          Hatches    drywell hatches or doors results Sealed(2)            pre-initiator                                  in a containment failure. The human error                                    1E-4 latent probability for each errors for failing                              of the three modeled hatches is to properly                                    non-significant compared to the close and test                                  CLERPs used in the analysis various DW                                      and non-significantly impacts hatches (DW                                    the calculated average personnel                                      SCLERP. A seismic event hatch, DW                                      does not impact the latent head and DW                                    probability of this postulated equipment                                      leakage.
hatch).
Each assigned 1E-4 probability (events 1DWHA-PERSON-F--;
1DWHA-DWHEAD-F--;
and 1DWHA-DWEQUIPF--).
Notes to Table E4-4:
(1)    This table lists the primary containment pathways modeled in the containment isolation fault tree logic of the Clinton Level 2 PRA. (Reference [35]).
(2)    The containment isolation fault tree also models contributions of pre-existing primary containment leakage and various containment hatches not properly sealed.
E4-31
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Summary Estimates of SCDF and SLERF have been derived for use in the CPS TSTF-505 program. Since the estimates are intended to be treated as conservative values in the RICT calculations for that program, the results for the case of plant level HCLPF = 0.30g PGA with c = 0.4 will be used, SCDF = 6.4E-6/yr and SLERF = 1.6E-6/yr.
For the Exelon BWR plants with a Mark I or II containment, if a RICT is being entered during a period when the containment is de-inerted, a different SLERF penalty (i.e., consistent with a CLERP = 1.0) may be applied to address the increased potential for hydrogen deflagration events given oxygen in the containment. This approach is not applicable for Clinton which has a different containment design and is not inerted with nitrogen. As the Clinton Mark III containment is not inerted during normal at-power operation (or at any time), hydrogen igniters in Mark III containments are used to burn hydrogen to prevent explosive levels of hydrogen developing post-core damage.
During postulated post-core damage scenarios without AC power the probability of hydrogen explosion induced containment failure increases significantly. This aspect is already included in the average SCLERP calculation (i.e., the SCLERP calculations include the increased probabilities of hydrogen explosion induced containment failure given no AC power to the igniters).
E4-32
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models 4    Extreme Winds Analysis This section provides an analysis of the High Winds / Tornados risk impact for Clinton Power Station (CPS). As described in Section 5 of the Clinton IPEEE (Reference [7]), the Clinton Power Station meets or exceeds the applicable Standard Review Plan sections and Regulatory Guides for high winds and tornadoes and concludes for the purpose of the IPEEE Generic Letter 88-20, Supplement 4, that the risk of wind or tornado is acceptably low. Various later high wind studies have been performed for Clinton and have concluded low risk from this hazard.
Wind Pressure As discussed in Section 3.3.2.1 of the Clinton USAR (Reference [15]), the wind loading design parameters of the Clinton design-basis tornado include:
* maximum wind speed of 230 mph
* maximum translational velocity of 46 mph
* external pressure drop of 1.2 psi Tornado wind speed hazard curve information for Clinton (as well as other U.S. nuclear plants) are provided in Table 6-1 of NUREG/CR-4461, Rev. 2 (Reference [39]). The NUREG/CR-4461 tornado hazard estimation methodologies include accepted practices and consider uncertainties. The Enhanced Fujita Scale based tornado hazard curve for the Clinton plant shows that the annual frequency of occurrence of the design-basis tornado on the Clinton site is <1E-7/yr. Table 6-1 of NUREG/CR-4461 provides a 1E-05, 1E-06 and 1E-07 annual exceedance frequency data point for each of the U.S. nuclear plant sites. The 1E-07 exceedance frequency of the Enhanced Fujita Scale hazard curve for the Clinton plant is 229 mph. Thus, even when not including the translational velocity, the Clinton design-basis tornado wind speed is below 1E-07/yr, using the Enhanced Fujita Scale.
Tropical storms (i.e., hurricanes) are not a concern at Clinton due its location (i.e., >600 miles inland from the coast). Straight winds (e.g., due to thunderstorms) are typically in the 50 - 70 mph range, although in rare cases may be over 100 mph. However, the hazard curve for straight winds tails off very quickly, such that below approximately 1E-3/yr, straight winds do not affect the overall wind hazard for areas with hurricane and/or tornado hazards (Reference [40]).
E4-33
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Tornado Missiles A number of tornado missile hazard analyses have been performed and updated for the Clinton plant. In 2007, an earlier Clinton TORMIS-based tornado missile risk analysis was updated to include the new Administration Building as an additional source of potential tornado-induced missiles (Reference [41]). TORMIS is a computer code based on EPRI methodologies used for assessing the probability of tornado-generated missiles hitting and damaging unprotected plant structures, systems and components (SSCs). The Reference [41] Clinton-specific TORMIS analysis was performed in accordance with the guidance described in the 1983 NRC TORMIS Safety Evaluation Report (Reference [42]). The results of the Clinton TORMIS studies are used in the Clinton USAR (Section 3.5) as risk informed justification for not requiring unique tornado missile protection barriers for certain components. The acceptance criterion used in the TORMIS study and the Clinton USAR follows US NRC Standard Review Plan Section 2.2.3, in that the total cumulative probability of a tornado-generated missile damaging an important system or component must be less that 1E-6/yr.
The Reference [41] Clinton 2007 TORMIS study base quantification exceeded the 1E-6/yr acceptance criterion. The study identified five (5) targets with the largest contribution to the result.
With those five targets missile-protected, the revised TORMIS analysis calculated the damage frequency for all SSCs unprotected against tornado missiles at 6.5E-7/yr (Table 8-9 (Reference [41]), which meets the 1E-6/yr SRP 2.2.3 acceptance criterion. The Clinton current plant configuration has since been revised such that these five targets are missile-protected. Per Clinton EC 366599 (Reference [43]), the following three targets were modified by installing missile shielding:
* VA system air intake in the north wall of Control Building EL. 825'
* VR system air intake in the north wall of Control Building EL. 825'
* Roll up door in the north wall of Control Building EL. 825' The remaining two targets (i.e., portions of the buried Shutdown Service Water (SX) system cable between the Screenhouse and the plant) were re-assessed and concluded the existing concrete enclosure and backfill provide adequate protection from the design missiles (no additional missile protection required). As such, the TORMIS-based calculation of the damage frequency for all SSCs unprotected against tornado missiles at the Clinton plant is 6.5E-7/yr.
E4-34
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Per SRP Section 2.2.3 requirements, Section 8 of the Clinton TORMIS study (Reference [41])
documents key conservatisms that exist in the analysis, including the following:
* Majority of the targets are building openings. These openings are not truly openings, instead these are penetrations for components such as piping etc. occupying the opening spaces. The resistance of these components to missile strike is ignored.
* Given that a missile has entered.an opening, there is an associated probability that it will not strike an actual safety-related component located inside the building. This probability depends upon the missile entry angle, missile size, location of safety-related components with respect to the opening and the size of the safety-related components. This probability is considered as unity for all openings. In other words, it has been assumed that any missile entering an opening will strike a safety-related component located inside the building.
* Given that a missile has entered an opening and has struck a safety-related component located inside the building, there is an associated probability that it will not damage the safety-related component. This probability depends upon the missile kinetic energy and load-displacement characteristics of the component. This probability also is considered as unity for all openings.
* Conservatism exists in the selection of the missile restraining force ratio range in the TORMIS model. The actual missile restraining force ratio varies between 5.9 to 423.2 for 12 missile types and their specific configurations. In the TORMIS model, the restraining force ratio range 10-15 was used. A relatively higher range based on statistical average (based missile population distribution) could have been used. This would reduce the target strike probability for the restrained missile case. It is also noted that in the restraint force calculation, the connection yield strength was used instead of the ultimate strength.
* Majority of the openings are 12" in. or less in diameter. The areas of all these openings were conservatively modeled as 5'x5'.
* The Clinton Unit 2 hole (elevation 698'-0") is located 39' below the average grade elevation of 737'-0". The base of the tornado was considered to be at the hole elevation of 698'-0" in the TORMIS model. This assumption implies maximum tornado wind speed at the average grade elevation, whereas near ground tornado wind speed is less than the maximum.
Later tornado wind/missile hazard studies were performed for Clinton in response to NRC RIS 2015-06, "Tornado Missile Protection" (Reference [44]), and NRC EGM 15-002, "Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance" (Reference [45]). The conclusion of these studies, per information in 2017 Clinton EC 620942 (Reference [46]), does not alter the conservative tornado-induced damage frequency calculated in the 2007 Clinton TORMIS study discussed above.
E4-35
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models High Winds Disposition for LAR Development for Risk Informed Applications The high wind hazard can be screened from consideration in this application as summarized below.
Wind Hazard Based on the plant design for wind pressure and the low frequency (<1E-7/yr) of design tornadoes, a demonstrably conservative estimate of CDF associated with high wind hazard (other than wind generated missiles) is much less than 1E-6/yr. Therefore, all non-missile high wind hazards can be screened from consideration based on EXT-C1 Criterion C of ASME/ANS RA-Sa-2009 (Reference
[3]).
Missile Hazard The TORMIS-based calculation of the damage frequency for all SSCs unprotected against tornado missiles at the Clinton plant is 6.5E-7/yr. The TORMIS analysis determines the total arithmetic sum of the tornado induced missile damage frequency for the identified unprotected SSCs, but the analysis does not specifically calculate core damage frequency (CDF) or large early release frequency (LERF). However, given the conservatism in TORMIS analyses and the fact that multiple targets must be failed in order to cause core damage, the CDF associated with tornado missiles is estimated to be much less than 1E-6/yr. Therefore, the CDF from the tornado missile hazard is estimated to be less than 1E-6/yr and high wind missile hazards can be screened from consideration based on EXT-C1 Criterion C of ASME/ANS RA-Sa-2009 (Reference [3]).
LERF Considerations There are no tornado missile targets associated with containment integrity, and the containment itself is extremely rugged with respect to tornado missiles. Therefore, Conditional Large Early Release Probability (CLERP) values should be similar to other weather-induced loss of offsite power sequences, such that no LERF-specific considerations are required in screening tornado missiles.
Configuration Specific Considerations An assessment of the high wind and tornado screening was performed considering SSCs out of service for maintenance. Based on the considerable missile protection and the limited vulnerabilities at Clinton, high wind and tornado risk would not be significantly affected by allowed maintenance configurations.
E4-36
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models 5    External Flooding Assessment Current Risk Basis The evaluation of the impact of the external flooding hazard at the site was updated as a result of the NRC's post-Fukushima 50.54(f) Request for Information. The station's flood hazard reevaluation report (FHRR) was submitted to NRC for review on March 12, 2014 (Reference [47]). The results indicate that all flood causing mechanisms, except Local Intense Precipitation (LIP), are bounded by the current licensing basis (CLB) and do not pose a challenge to the plant based on its design. The NRC staff issued a staff assessment concluding that no further action is required to respond to the 50.54(f) RFI on September 3, 2015 (Reference [48]) and the plant response is considered adequate to respond to an external flood event.
However, the calculations used to determine LIP water surface elevations (WSEs) were found to have an error in rain-on-building applications. The condition was entered into the plants corrective action program (CAP) and documented in Issue Report (IR) No. 2406577. The model used to perform the calculations was corrected in calculation IP-S-0282 and resubmitted to NRC for Review (Reference [49]). It was determined that two external doors to the Radwaste Building would be subjected to approximately 1.2 inches of standing water. The calculation assumed that minimal to no leakage is anticipated through the airlock door (1DR1-327) due to its construction and opening procedure that prevents both doors from opening at the same time. In leakage is expected through the north east exterior rolling door (1SD1-49). The CPS Mitigating Strategies Assessment (MSA) concluded there are no safety related (SR) SSCs in the Radwaste (Reference [50]) and therefore, no impacts to SR functions due to the LIP flood causing mechanism. Additionally, there are no manual actions required by station personnel in anticipation of any external flooding event.
Challenges Posed Weather-induced Loss of Offsite Power (LOSP) is a potential challenge.
Disposition for RICT Program:
All non-LIP external flooding mechanisms were considered bounded by the plant's CDB. These mechanisms will not produce external flooding that will challenge any SR SSCs relied upon to safely shutdown the plant.
External flooding from LIP will similarly not challenge any safety functions at CPS. The max WSE is calculated to be 1.2-inches above the building finished floor elevation at the Radwaste Building. In calculation IP-S-0282 and the MSA (Reference [50]), it is shown that there are no SR SSCs in the Radwaste Building and water will not propagate or accumulate in any other buildings containing SR SSCs.
E4-37
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Therefore, all external flooding mechanisms are screened from inclusion in the RICT calculations because the probable maximum floods from each mechanism calculated at CPS will not cause any adverse impacts to safety functions. The external flooding event damage potential is less than that for which the plant is designed and can be screened based on EXT-B1 Criterion C1 of ASME/ANS RA Sa 2009 (Reference [3]).
Configuration Specific Considerations There are no configuration specific considerations related to the screening assessment provided above for CPS.
E4-38
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models 6    Evaluation of External Event Challenges and IPEEE Update Results This Section provides an evaluation of other external hazards. The results of the assessment of these hazards is provided in Table E4-7. Table E4-8 provides the summary criteria for screening of the hazards listed in Table E4-7.
Hazard Screening The IPEEE for CPS provides an assessment of the risk to CPS associated with other external hazards.
Additional analyses have been performed since the IPEEE to provide updated risk assessments of various hazards, such as aircraft impacts, industrial facilities and pipelines, and external flooding.
These analyses are documented in the UFSAR.
Table E4-7 reviews and provides the bases for the screening of external hazards, identifies any challenges posed, and identifies any additional treatment of these challenges, if required. The conclusions of the assessment, as documented in Table E4-7, assure that the hazard either does not present a design-basis challenge to CPS or is adequately addressed in the PRA.
Impacts to RICT In the application of Risk-Informed Completion Times, 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 CPS has been performed accounting for such configuration-specific impacts. This evaluation involves several steps.
As a first step in this screening process, hazards that screen for one or more of the following criteria (as defined in Table E4-8) still screen regardless of the configuration, as these criteria are not dependent on the plant configuration.
* The occurrence of the event is of sufficiently low frequency that its impact on plant risk does not appreciably impact CDF or LERF. (Criterion C2)
* The event cannot occur close enough to the plant to affect it. (Criterion C3)
* The event which subsumes the external hazard is still applicable and bounds the hazard for other configurations (Criterion C4)
* The event develops slowly, allowing adequate time to eliminate or mitigate the hazard or its impact on the plant. (Criterion C5)
E4-39
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models The next step in the screening process is to consider the remaining hazards (i.e., those not screened per the above criteria) to consider the impact of the hazard on the plant given particular configurations for which a RICT is allowed. For hazards for which the ability to achieve safe shutdown may be impacted by one or more such plant configurations, the impact of the hazard to particular SSCs is assessed and a basis for the screening decision applicable to configurations impacting those SSCs is provided.
As noted above, the configurations to be evaluated are those involving unavailable SSCs whose LCOs are included in the RICT program.
Evaluation of Seismic Induced Loss of Offsite Power Past TSTF-505 applications have also included evaluation of any incremental risk associated with challenges to the facility that do not exceed the design capacity and the past submittals have focused on the challenge of seismically induced LOOP. The methodology for computing the seismically-induced LOOP frequency is to convolve the CPS mean seismic hazard curve with the offsite power fragility and the past TSTF-505 applications have approached this discussion conservatively by performing the convolution over the entire hazard curve (not just below the design basis). That same approach is used in this discussion. The CPS seismic hazard curve is as described previously in Table E4-1.
Table E4-5 provides the mean seismic hazard data and the LOOP seismic-induced failure probability (increasing with increase seismic magnitude) based on the fragility of offsite power. The convolution calculation includes the entire hazard curve from earthquakes magnitudes well below the CPS operating basis earthquake (OBE=0.11g) to well beyond the CPS safe shutdown earthquake (SSE=0.26g).
The failure probabilities for LOOP are represented by failure of ceramic insulators in the power distribution system, based on the following fragility data from Table A-0-4 of the NRC RASP Handbook, Volume 2 (Reference [51]), this is a common offsite power fragility used for Central and Eastern US SPRAs:
Offsite Power Capacity (ceramic insulators): Am = 0.30g; R = 0.30, U = 0.45 Given the mean frequency and failure probability for each seismic interval, it is straightforward to compute the estimated frequency of seismically induced loss of offsite power for the CPS site by taking the product of the interval frequency and the offsite power failure probability. As shown in Table E4-5, the total seismic LOOP frequency across the entire seismic hazard curve approximately 9.7E-5/yr. Note that this overstates the below-design challenge frequency but is conservative for this purpose.
E4-40
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models The FPIE PRA models LOOP from plant-centered, switchyard-centered, grid-related, and weather-related events. Based on the CPS FPIE PRA, the total frequency of unrecovered loss of offsite power (i.e., the sum of the frequency multiplied by the non-recovery probability at 24 hours over these LOOP events), is 1.7E-3/yr, as shown in Table E4-6.
The total (i.e., across the entire hazard curve) seismically-induced (unrecoverable) LOOP frequency is approximately 6% of the total unrecovered LOOP frequency already addressed in the PRA. The below-design basis (i.e., less than SSE of 0.26g) seismic-induced LOOP frequency is approximately 4% of the total unrecovered LOOP frequency already addressed in the PRA; this frequency is judged to be a reasonably small fraction that it will not significantly impact the RICT Program calculations and it can be omitted.
E4-41
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4 Convolution Summary of CPS Seismic CDF CL Offsite Power HCLPF                                              Convolution Calculation CL Seismic Hazard Curve (Reference 51)                                  (CL Offsite Power HCLPF fragility with Seismic Hazard)
Hazard Interval                Hazard Mean      Representative Hazard          Interval Peak Ground Exceedance        Magnitude      Interval  Occurrence    Convolved (1)
HCLPF        Am            Acceleration Frequency    (geo. mean, g Fragility    Frequency      Frequency (g, PGA) (g, PGA)    c        (g)        (/yr)        PGA)        (Mean)        (/yr)          (/yr) 0.09      0.30    0.54      0.0005    9.59E-02      7.07E-04    1.89E-29      1.45E-02      2.73E-31 0.001      8.14E-02      2.24E-03    5.52E-20      4.76E-02      2.63E-21 0.005      3.38E-02      7.07E-03    1.83E-12      1.51E-02      2.77E-14 0.01      1.87E-02      1.22E-02    1.48E-09      6.30E-03      9.33E-12 0.015      1.24E-02      2.12E-02    4.37E-07      7.25E-03      3.17E-09 0.03      5.15E-03      3.87E-02    7.09E-05      2.95E-03      2.09E-07 0.05      2.20E-03      6.12E-02    1.55E-03      1.21E-03      1.87E-06 0.075      9.94E-04      8.66E-02    1.02E-02      4.54E-04      4.65E-06 0.1      5.40E-04      1.22E-01    4.68E-02      3.24E-04      1.52E-05 0.15      2.16E-04      2.12E-01    2.54E-01      1.75E-04      4.46E-05 0.3      4.09E-05      3.87E-01    6.75E-01      2.97E-05      2.00E-05 0.5      1.12E-05      6.12E-01    9.03E-01      7.49E-06      6.76E-06 0.75      3.71E-06      8.66E-01    9.74E-01      2.10E-06      2.05E-06 1      1.61E-06      1.22E+00    9.95E-01      1.15E-06      1.14E-06 1.5      4.61E-07      2.12E+00    1.00E+00 4.18E-07            4.18E-07 3      4.30E-08      3.87E+00    1.00E+00 3.71E-08            3.71E-08 5      5.92E-09      6.12E+00    1.00E+00 4.90E-09            4.90E-09 7.5      1.02E-09      8.66E+00    1.00E+00 7.55E-10            7.55E-10 10      2.65E-10      1.00E+01    1.00E+00 2.65E-10            2.65E-10 Total Convolved Seismic LOOP Across Hazard Curve (1/yr):              9.7E-05 E4-42
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-6 LOSS OF OFFSITE POWER (LOOP) NON-RECOVERY FREQUENCY CPS FPIE PRA Probability of      24-hr Non-LOOP Initiator Non-Recovery of    Recovered LOOP Contributor        Contributor Offsite AC by 24    LOOP Frequency)
Hrs                Frequency (/yr)
(/yr)
Plant-Centered          1.99E-03            8.14E-03            1.6E-05 Switchyard-Centered    1.34E-02            2.25E-02            2.9E-04 Grid-Related            7.43E-03            2.11E-02            1.5E-04 Weather-Related        5.97E-03            2.25E-01            1.3E-03 Total Random Unrecovered LOOP Frequency                          1.7E-03 E4-43
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                Definition                              Criterion                          CPS Response (Y/N)
(Note a)
In the IPEEE (Reference [7]), a probabilistic bounding analysis was performed for aircraft. The median frequency of aircraft accidents which could lead to radiological consequences in excess of the exposure guidelines of 10 CFR Part 100 was calculated less than 1E-7/year (PS2).
Sections 2.2.2.5 and 3.5.1.6 of the USAR (Reference [15]) describe the airports and airways in the vicinity of the site.
A direct or indirect (i.e.,                                    a. There is one federal Low Altitude Federal Airway skidding impact) collision of a                                      with its centerline passing within 2 miles east of the PS2 portion of or an entire aircraft                                    station. Three additional Low Altitude Federal Aircraft impacts                                          Y with one or more structures at                                      Airways within 6 miles were evaluated. The or in the area surrounding the                        PS4            calculated frequency of aircraft is 0.54E-7/year plant site.                                                          (PS4).
: b. There are no commercial airports within 10 miles of the site.
: c. There are three private airstrips within 5 miles of the station.
Based on this review, the Aircraft impact hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
E4-44
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                Definition                              Criterion                        CPS Response (Y/N)
(Note a)
The mid-western location of CPS station precludes the possibility of an avalanche.
A rapid flow of a large mass of Based on this review, the Avalanche hazard can be accumulated frozen Avalanche                                                  Y            C3      considered to be negligible.
precipitation down a sloped surface.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Hazard is slow to develop and can be identified via monitoring and managed via standard maintenance process. Actions committed to and completed by CPS in response to Generic Letter 89-13 provide on-going The accumulation or                                            control of biological hazards. These controls are deposition of vegetation or                                    described in Exelon procedure ER-AA-340, "GL 89-13 organisms (e.g., zebra                                          Program Implementing Procedure" (Reference [53]).
Biological events mussels, clams, fish) on an              Y            C5 intake structure or internal to a Based on this review, the Biological Event hazard can system that uses an intake be considered to be negligible.
structure.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
E4-45
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                Definition                            Criterion                        CPS Response (Y/N)
(Note a)
The mid-western location of CPS station precludes the possibility of coastal erosion.
The wearing away of a Based on this review, the Coastal Erosion hazard can shoreline due to wave action, Coastal erosion                                          Y            C3      be considered to be negligible.
tidal currents, wave currents, drainage, or winds.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Drought is a slowly developing hazard allowing time for orderly plant reductions, including shutdowns.
An extended period of months or years when a region                                        Based on this review, the Drought hazard can be Drought          experiences a deficiency in its        Y            C5      considered to be negligible.
surface or underground water supply.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Flooding that results from                                    See Section 5 of this enclosure for results and water sources external to the                                justification of screening of external flooding related External Flooding plant other than those explicitly      Y            C1      hazards.
identified elsewhere in this table.
E4-46
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                    Definition                            Criterion                          CPS Response (Y/N)
(Note a)
Extreme wind from straight                                    See Section 4 of this enclosure for results and wind sources (e.g.,                                  PS2      justification of screening of extreme wind or tornado Extreme Wind or thunderstorms and                      Y                      hazards.
Tornado extratropical cyclones) and                          PS4 tornados.
Fog is discussed in the UFSAR Section 2.3 (Reference [15]).
The principal effects of such events (such as freezing fog) would be to cause a loss of off-site power which is addressed in the weather-related Loss of Offsite Power Water droplets suspended in                                    initiating event in the internal events PRA model for the atmosphere at or near the                                  CPS.
Fog                                                          Y            C4 Earth's surface that limit visibility.                                                    Based on this review, the Fog hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Forest fires and grass fires were screened in the IPEEE Fires originating from outside (Reference [7]). FSAR Section 2.2.3.1.4 also the plant site boundary that discusses this hazard and states that forest or brush Forest or Range Fire are caused by the uncontrolled          Y            C3 fires cannot pose any danger because of the site combustion of vegetation (e.g.,
landscaping.
trees, grasses, brush, etc.).
E4-47
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard            Definition                              Criterion                          CPS Response (Y/N)
(Note a)
Based on this review, the Forest or Range Fire hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Frost is discussed in the UFSAR Section 2.5 (Reference [15]). The principal effects of such events would be to cause a loss of off-site power and are addressed in the weather-related Loss of Offsite Power A thin layer of ice crystals that                              initiating event in the internal events PRA model for form on the ground or the                                      CPS.
surface of an earthbound Frost                                                Y            C4 object when the temperature Based on this review, the Frost hazard can be of the ground or surface of the considered to be negligible.
object falls below freezing.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Per the IPEEE (Reference [7]), hailstorm need not be considered per the guidance contained in NUREG Showery precipitation in the                                    1407 (Reference [6]).
Hail        form of irregular pellets or balls      Y            C4      The principal effects of such events would be to cause of ice.                                                        a loss of off-site power and are addressed in the weather-related Loss of Offsite Power initiating event in the internal events PRA model for CPS.
E4-48
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard          Definition                          Criterion                        CPS Response (Y/N)
(Note a)
Based on this review, the Hail hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
The principal effects of such events would result in elevated lake temperatures which are monitored during performance of control room shiftly checks. Should the ultimate heat sink temperature reach 93&deg;F then operations' procedures require further increased monitoring and development of compensatory measures.
This phenomenon provides large amount of time for High summer  High abnormal ambient                                    preparation (weather forecast) with time for Y            C5 temperature  temperatures.                                            implementation of appropriate mitigation actions.
Based on this review, the High Summer Temperature hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
E4-49
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                      Definition                            Criterion                          CPS Response (Y/N)
(Note a)
High tide or river stage does not apply since there is only a cooling lake (Lake Clinton).
Per USAR Section 2.4, the cooling lake (Lake Clinton) was formed by a dam with spillways to control high lake level (C1). In addition, the event develops slowly, allowing adequate time to eliminate or mitigate the The periodic maximum rise of threat (C5).
sea level resulting from the C1 High tide, Lake      combined effects of the tidal Y
Level, or River Stage gravitational forces exerted by                                See also "External Flooding."
the Moon and Sun and the                              C5 rotation of the Earth.
Based on this review, the High Tide, Lake Level, or River Stage hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
The mid-western location of CPS precludes the possibility of a hurricane.
An extremely large, powerful, and destructive storm resulting                                Based on this review, the Hurricane hazard can be Hurricane            in strong winds, excessive              Y            C3      considered to be negligible.
rainfall, high waves, storm surge, and tornados.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
E4-50
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                    Definition                              Criterion                        CPS Response (Y/N)
(Note a)
Per the IPEEE (Reference [7]), ice storm need not be considered per the guidance contained in NUREG 1407 (Reference [6]).
The principal effects of such events would be to cause a loss of off-site power and are addressed in the The accumulation of frozen                                      weather-related Loss of Offsite Power initiating events water on bodies of water (e.g.,                                  in the internal events PRA model for CPS.
Ice cover              lakes, rivers, etc.) or on                Y            C4 structures, systems, and components.                                                      Based on this review, the Ice Cover hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Per FSAR Section 2.2.1 (Reference [15]), there are several military reserve unit armories located in the An accident at an offsite general area of the site which are listed in Table 2.2-1.
industrial or military facility The closest is the Bloomington armory located 23 miles such as a release of toxic C1      NNW. The armories normally should contain no Industrial or military gases, a release of Y                      explosives. There are no military missile sites within 50 facility accident      combustion products, a miles of the station (C3).
release of radioactivity, an                            C3 Per FSAR Section 2.2.3 (Reference [15]), no nearby explosion, or the generation of industrial or other activities have been identified which missiles.
could pose unusual hazards to the Clinton Power Station.
E4-51
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                Definition                            Criterion                          CPS Response (Y/N)
(Note a)
All hazardous materials stored or shipped in the vicinity of CPS were evaluated in FSAR Subsection 2.2.3.1.3 for their toxic potential on control room habitability.
Based on this evaluation, releases of hazardous materials in the vicinity of CPS are not considered as design basis accidents (C1).
See also "Toxic Gas."
Based on this review, the Industrial or Military Facility Accident hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Flooding from sources internal                                The CPS Internal Events PRA includes evaluation of Internal Flooding                                        N/A          N/A to the plant.                                                  risk from internal flooding events.
Fires originating from inside                                  The CPS Internal Fire PRA includes evaluation of risk the plant site boundary that                                  from internal fire events.
Internal Fire    are caused by, for example,            N/A          N/A combustible material or electrical shorts.
E4-52
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard            Definition                              Criterion                        CPS Response (Y/N)
(Note a)
The mid-western location of CPS precludes the possibility of a landslide.
A rapid flow of a large mass of earth, rock, or material other                                Based on this review, the Landslide hazard can be Landslide    than accumulated frozen                Y            C3      considered to be negligible.
precipitation down a sloped surface.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Per the IPEEE (Reference [7]), the design of CPS includes features to protect against lightning (C1).
Lightning strikes are not uncommon in nuclear plant experience. They can result in losses of off-site power or surges in instrumentation output if grounding is not fully effective. The latter events often lead to reactor An electrical discharge from a                        C1      trips. Both events are incorporated into the CPS Lightning    cloud to the ground or                  Y                      internal events model through the incorporation of Earthbound object.                                  C4      generic and plant specific data (C4).
Based on this review, the Lightning hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
E4-53
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                Definition                              Criterion                        CPS Response (Y/N)
(Note a)
Per USAR Section 2.4, the cooling lake (Lake Clinton) was formed by a dam with outlet works provided to control low lake level (C1). In addition, the effect of low lake level would take place slowly allowing time for orderly plant reductions, including shutdowns (C5).
A decrease in the water level                        C1 Low Lake Level or of the lake or river used for          Y River Stage                                                                      Based on this review, the Low Lake Level or River power generation.                                    C5      Stage hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Per the IPEEE (Reference [7]), low winter temperature need not be considered per the guidance contained in NUREG 1407 (Reference [6]). However, there are existing severe weather procedures and cold weather checklists that are performed in advance of the onset of cold weather to allow adequate time to eliminate or Low winter        Low abnormal ambient                                          mitigate the threat (C5).
Y            C5 temperature      temperatures.
Based on this review, the Low Winter Temperature hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
E4-54
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                      Definition                              Criterion                        CPS Response (Y/N)
(Note a)
Per the IPEEE (Reference [7]), meteorite or satellite need not be considered per the guidance contained in NUREG 1407 (Reference [6]). However, the frequency A meteoroid or artificial                                      of a meteor or satellite strike is judged to be so low as satellite that releases energy                                  make the risk from such events insignificant.
Meteorite or Satellite due to its disintegration in the atmosphere above the Earth's            Y            PS4 Impact                                                                                Based on this review, the Meteorite or Satellite hazard surface, direct impact with the Earth's surface, or a                                          can be considered to be negligible.
combination of these effects.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Per USAR Table 2.2-4 (Reference [15]) there is one pipeline within 5 miles of the site (at about 4650 feet) that transports refined petroleum products. Per USAR Section 2.2.3.1.1 the distance of 4600 feet has been established as a limit beyond which a possible pipeline An accident involving the rupture followed by an explosion under conservative rupture of a pipeline carrying Pipeline accident                                              Y            C1      weather conditions does not govern the design of the hazardous materials or toxic plant. Since the pipelines that existed prior to gases.
construction of the plant have been relocated (USAR Section 2.2.2.3) and the closest pipeline passes about 4650 feet from the site, explosions do not pose any hazard to the plant.
E4-55
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                  Definition                              Criterion                        CPS Response (Y/N)
(Note a)
Based on this review, the Pipeline Accident hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Compliance with Regulatory Guide 1.78 for hazardous chemicals stored onsite is described in USAR Section 6.4 (Reference [15]). Gaseous chlorine is no longer allowed on site by plant procedure and there are no other significant depots of chlorine within a five mile An onsite accident involving radius of the site. Of the other potentially hazardous the storage or handling of chemicals stored on site, listed in USAR Table 2.2-6, hazardous materials such as a only sulfuric acid, carbon dioxide, and nitrogen are release of toxic gases, a included in Regulatory Guide 1.78.
release of combustion Release of                                                                        The following are features protecting against potential products, a release of Chemicals in Onsite                                        Y            C1      problems upon a release of sulfuric acid:
radioactivity, an explosion, or Storage the generation of missiles. In                                    a. Sulfuric acid has a low vapor pressure (< 1 Torr),
this context, an onsite release                                  b. The relative location of the sulfuric acid storage of radioactivity is assumed to                                    facility with respect to the control room minimum be associated with lowlevel                                      outside air intakes, and radioactive waste.
: c. The acid storage tank is vented to the outside.
Fumes from spillage within the acid storage area are diluted by the exhaust air from the sulfuric acid storage area with the radwaste building and balance of the plant exhaust air streams.
E4-56
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard              Definition                              Criterion                        CPS Response (Y/N)
(Note a)
Analysis has shown that a postulated rupture in the carbon dioxide storage system does not result in an unacceptable concentration of CO2 within the control room. Since the amount of nitrogen stored onsite is not a significant fraction of the control room volume, per Regulatory Guide 1.78, it does not need to be considered.
Based on this review, the Release of Chemicals in Onsite Storage hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Per UFSAR 2.4.1.2 (Reference [15]), the Salt Creek River is the principal tributary of the Sangamon River, The redirection of all or a                                    which drains into the Illinois River. Per UFSAR 2.4.9, portion of river flow by natural                                there is no historical evidence of channel diversion of causes (e.g., a riverine                                        Salt Creek and North Fork of Salt Creek upstream of River diversion                                          Y            C3      the dam site.
embankment landslide) or intentionally (e.g., power production, irrigation, etc.).                                  Based on this review, the River Diversion hazard can be considered to be negligible.
E4-57
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                  Definition                              Criterion                        CPS Response (Y/N)
(Note a)
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Per the IPEEE (Reference [7]), sand or dust storm need not be considered per the guidance contained in NUREG 1407 (Reference [6]).
A strong wind storm with Based on this review, the Sand or Dust Storm hazard Sand or Dust Storm airborne particles of sand and          Y            C1 can be considered to be negligible.
dust.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Per USAR Section 2.4.5, there is no large body of water near the site where significant seiche formations can occur. The size of the cooling lake is not large enough to develop a seiche flooding condition which is An oscillation of the surface of                              more critical than the probable maximum flood (PMF) a landlocked body of water,                                    condition.
Seiche            such as a lake, that can vary in        Y            C1 period from minutes to several See also "External Flooding".
hours.
Based on this review, the Seiche hazard can be considered to be negligible.
E4-58
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                Definition                              Criterion                        CPS Response (Y/N)
(Note a)
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
A sudden release of energy                                      The risk contribution from seismic events is evaluated Seismic activity  from the Earth's crust resulting        N            N/A      quantitatively. See Section 3 for results of the plant-in strong ground motion.                                        specific analysis of seismic risk to CPS.
Per the IPEEE (Reference [7]), snow storm need not be considered per the guidance contained in NUREG 1407 (Reference [6]).
The accumulation of snow on Based on this review, the Snow hazard can be Snow              structures, systems, and                Y            C1 considered to be negligible.
components.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
USAR Chapter 2 (Reference [15]) discusses site characteristics and stability of soils. Extensive geotechnical investigations carried out prior to and The relative change in volume during construction (including geologic mapping of the of the soil as a result of the                        C1 Soil shrink-swell                                          Y                      excavations) showed nothing that would preclude safe type of soil and the amount of construction or operation of a nuclear-fueled power moisture.
station. There are no known faults or folds of design significance at or anywhere near the site.
E4-59
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard            Definition                            Criterion                        CPS Response (Y/N)
(Note a)
Based on this review, the Soil Shrink-Swell Consolidation hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
An abnormal rise in sea level                                The mid-western location of CPS precludes the accompanying a hurricane or                                  possibility of a sea level driven storm surge.
other intense storm, whose height is the difference                                    Based on this review, the Storm Surge hazard can be Storm surge  between the observed level of        Y            C3      considered to be negligible.
the sea surface and the level that would have occurred in There are no configuration-specific considerations for the absence of the intense this hazard. This hazard can be excluded from the storm.
RICT program evaluation.
E4-60
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard            Definition                              Criterion                        CPS Response (Y/N)
(Note a)
UFSAR Section 2.2.3.1.3 (Reference [15]) discusses toxic gas.
Van Horn - DeWitt is the only facility within five miles of the site which manufactures, uses, or stores toxic chemicals. Van Horn - DeWitt is a distributor of agricultural products and chemicals (such as Release of gas in proximity to                                pesticides, herbicides, and fertilizers) and their facility site that can impact health and                                in DeWitt is located approximately 2.5 miles from safety of plant personnel. An                                  Clinton Power Station.
onsite accident involving the                                  CPS reviewed a list of chemicals distributed by Van storage or handling of                                        Horn - DeWitt, and determined that with the exception hazardous materials such as a                                  of anhydrous ammonia, none of the chemicals require release of toxic gases, a                            C1      evaluation for their potential effect on control room Toxic Gas    release of combustion                  Y                      habitability (due to an accidental spill or release) in products, a release of                                C3      accordance with Regulatory Guide 1.78. Calculations radioactivity, an explosion, or                                (Reference [54] show the postulated accidents of the the generation of missiles. In                                anhydrous ammonia nurse tanks and tanker trucks this context, an onsite release                                used by farmers and suppliers do not adversely affect of radioactivity is assumed to                                the control room habitability (C1).
be associated with lowlevel                                  Reference [55] concluded that all the identified toxic radioactive waste.                                            chemicals (transported via roadways) do not need further evaluation.
In addition, Per FSAR Section 6.4.4.2, gaseous chlorine is no longer allowed on site by plant procedure and there are no other significant depots of chlorine within a five mile radius of the site. Therefore, no automatic initiation of the control room ventilation E4-61
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard              Definition                              Criterion                          CPS Response (Y/N)
(Note a) chlorine mode and no chlorine detectors are required (C3).
See also Release of Chemicals in onsite storage, industrial or military facility accident and transportation Accidents.
Based on this review, the Toxic Gas hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
Transportation accidents was evaluated in the IPEEE (Reference [7]) and in USAR Section 2.2 (Reference [15]). In the IPEEE, the evaluation was conducted against the NRC Standard Review Plan An accident involving damage which concluded that the risk was acceptably low to a landbased or marine (PS2).
vehicle transporting hazardous                        C1 Transportation                                                                Per FSAR 2.2.3 (Reference [15]), no nearby industrial materials that may result in a          Y accidents                                                                    or other activities have been identified which could release of toxic gases, a                            PS2 release of combustion                                          pose unusual hazards to the Clinton Power Station.
products, or an explosion.                                    The nearest highway is State Highway 54 which passes about three-quarters of a mile from the reactor containment building. U.S. Highway 51, is approximately 6 miles from the site. The nearest railroad is the Gilman Line of the Canadian E4-62
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard            Definition                            Criterion                        CPS Response (Y/N)
(Note a)
National/Illinois Central Railroad which runs parallel to Highway 54 and traverses north of the site approximately .75 miles. Effects of accidents on these transportation routes have been evaluated and it is concluded that they need not be considered as design basis events. The station is not located near a navigable waterway (C1).
Based on this review, the Transportation Accident hazard can be considered to be negligible.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
The mid-western location of CPS precludes the possibility of a tsunami.
A sea wave of local or distant origin that results from Based on this review, the Tsunami hazard can be largescale seafloor Tsunami                                            Y            C3      considered to be negligible.
displacements associated with large earthquakes or major submarine slides or landslides.                              There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
E4-63
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                  Definition                              Criterion                        CPS Response (Y/N)
(Note a)
Turbine generated missiles are discussed in UFSAR Section 3.5.1.3 (Reference [15]). With the replacement The generation of a                                              of the Low Pressure (LP) rotors, all the turbine rotors highenergy missile that is                                      are of the monoblock design. The monoblock rotors ejected from the turbine casing                                  have very low stress level. Missile generation due to resulting from failure of a                                      turbine failure is generally postulated to be caused by steam turbine. The                                              turbine overspeed. General Electric has established turbinegenerated missile may                                    that the speed capability of these rotors is considerably be ejected either upward (i.e.,                                  higher than the maximum attainable speed of these Turbine-generated hightrajectory missile) which            Y            PS4      turbine generator units. Consequently, the probability missiles may result in damage to                                          of missiles being generated is statistically insignificant.
safetyrelated structures, systems, and components Based on this review, the Turbine-Generated Missiles (SSCs) from the falling missile hazard can be considered to be negligible.
or it may be ejected directly toward safetyrelated SSCs (i.e., lowtrajectory missiles).                                There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
E4-64
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4-7 Evaluation of Other External Hazards Screening Screened Hazard                Definition                              Criterion                        CPS Response (Y/N)
(Note a)
The extrusion of magma from                                    Not applicable to the site because of location (no active beneath the earth's crust that                                or dormant volcanoes located near plant site).
may be accompanied by the flow of lava and explosion of Based on this review, the Volcanic Activity hazard can fragmented material Volcanic activity                                        Y            C3      be considered to be negligible.
(pulverized pieces of rock, bits of chilled magma), and releases of volcanic ash and                                  There are no configuration-specific considerations for dust as well as gases and                                      this hazard. This hazard can be excluded from the steam.                                                        RICT program evaluation.
Waves associated with adjacent large bodies of water are not applicable to the site. Waves associated with external flooding are covered under that hazard.
An area of moving water that is raised above the main                                      See also External Flooding surface of an ocean, a lake, Waves                                                    Y            C1 etc. as a result of the wind                                  Based on this review, the Waves hazard can be blowing over an area of fluid                                  considered to be negligible.
surface.
There are no configuration-specific considerations for this hazard. This hazard can be excluded from the RICT program evaluation.
E4-65
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4- 8: Progressive Screening Approach for Addressing External Hazards Event Analysis                        Criterion                    Source      Comments NUREG/CR-2300 C1. Event damage potential is <        and ASME/ANS events for which plant is designed. Standard RA-Sa-2009 NUREG/CR-2300 C2. Event has lower mean frequency and ASME/ANS and no worse consequences than Standard RA-Sa-other events analyzed.
2009 NUREG/CR-2300 Initial Preliminary    C3. Event cannot occur close enough    and ASME/ANS Screening        to the plant to affect it.              Standard RA-Sa-2009 Not used to NUREG/CR-2300 screen. Used C4. Event is included in the definition and ASME/ANS only to include of another event.                      Standard RA-Sa-within another 2009 event.
C5. Event develops slowly, allowing    ASME/ANS adequate time to eliminate or mitigate  Standard RA-Sa-the threat.                            2009 ASME/ANS PS1. Design-basis hazard cannot Standard RA-Sa-cause a core damage accident.
2009 NUREG-1407 and PS2. Design basis for the event ASME/ANS meets the criteria in the NRC 1975 Standard RA-Sa-Standard Review Plan (SRP).
2009 Progressive                                                NUREG-1407 as Screening        PS3. Design-basis event mean modified in frequency is < 1E-5/y and the mean ASME/ANS conditional core damage probability is Standard RA-Sa-
                      < 0. 1.
2009 NUREG-1407 and PS4. Bounding mean CDF is < 1E-        ASME/ANS 6/y.                                    Standard RA-Sa-2009 E4-66
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models Table E4- 8: Progressive Screening Approach for Addressing External Hazards Event Analysis                    Criterion                Source          Comments NUREG-1407 and Screening not successful. PRA ASME/ANS Detailed PRA      needs to meet requirements in the Standard RA-Sa-ASME/ANS PRA Standard.
2009 E4-67
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models 7    Conclusions Based on this analysis of external hazards for CPS, no additional external hazards other than seismic events need to be added to the existing PRA model. CPS will apply a seismic "penalty" in the risk evaluations performed as part of the process to calculate a Risk Informed Completion Time (RICT).
The evaluation concluded that all other hazards either do not present a design-basis challenge to CPS, the challenge is adequately addressed in the PRA, or the hazard has a negligible impact on the calculated RICT and can be excluded.
The ICDP/ILERP acceptance criteria of 1E-5/1E-6 will be used within the PARAGON framework to calculate the resulting RICT and RMAT based on the total configuration-specific delta CDF/LERF attributed to internal events and internal fire, plus the seismic delta CDF/LERF penalty factors.
E4-68
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models 8    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, October 12, 2012 (ADAMS Accession No. ML12286A322).
[2]  Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"
May 17, 2007 (ADAMS Accession No. ML071200238).
[3]  ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum A to RAS-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.
[4]  NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk- Informed Decision Making," Revision 1, March 2017.
[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]  Clinton Power Station Individual Plant Examination for External Events Final Report, September 1995.
[8]  Lettis Consultants International (LCI), Inc., Project No. 1041, "Clinton Seismic Hazard and Screening Report," December 18, 2013.
[9]  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 NTTF Review of the Fukushima Dai-ichi Accident," Clinton Power Station Unit 1, March 31, 2014 (ML14091A011).
[10] Electric Power Research Institute (EPRI) 3002000709, "Seismic Probabilistic Risk Assessment Implementation Guide," December 2013.
E4-69
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
[11] Generic Issue (GI) 199, "Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants," U.S. NRC Information Notice (IN) 2010-18, September 2, 2010; Tables B.2, C.1 and C-2.
[12] Electric Power Research Institute (EPRI) NP-6041-SL, "A Methodology for Assessment of Nuclear Plant Seismic Margin," Revision 1, August 1991.
[13] NUREG/CR-0098, "Development of Criteria for Seismic Review of Selected Nuclear Power Plants," May 1978.
[14] Regulatory Guide 1.60, "Design Response Spectra for Seismic Design of Nuclear Power Plants," December 1973.
[15] Clinton Power Station Updated Safety Analysis Report (USAR), Revision 20, October 2018.
[16] CPS Full-Power Internal Event PRA Model of Record (MOR), CL117B.
[17] Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specifications to Implement NEI 06-09, Revision 0, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"
(Enclosure E3), September 13, 2012, NRC ADAMS Accession # ML12258A055.
[18] Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 1, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b,"
February 25, 2016, NRC ADAMS Accession # ML16060A223.
[19] CL-MISC-026, "External Hazards Assessment for CPS," Revision 0, March 2020.
[20] Electric Power Research Institute (EPRI) 1025287, "Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic," February 2013.
[21] Beaver Valley Power Station Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1, July 2017, NRC ADAMS Accession No. ML17213A017.
E4-70
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
[22] Browns Ferry Nuclear Plant Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1, December 2019, NRC ADAMS Accession No. ML19351E391.
[23] Callaway Plant Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1, August 2019, NRC ADAMS Accession No. ML19225D324.
[24] Columbia Generating Station Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1, September 2019, NRC ADAMS Accession No. ML19273A907.
[25] DC Cook Nuclear Plant Seismic Probabilistic Risk Assessment in Response to 50.54(f)
Letter with Regard to NTTF 2.1, November 2019, NRC ADAMS Accession No. ML19310D805.
[26] Diablo Canyon Plant Seismic Hazard and Screening Report in Response to 50.54(f)
Letter with Regard to NTTF 2.1, March 2015, NRC ADAMS Accession No. ML15070A607.
[27] Diablo Canyon Plant Seismic Probabilistic Risk Assessment in Response to 50.54(f)
Letter with Regard to NTTF 2.1, April 2018, NRC ADAMS Accession No. ML18120A201.
[28] North Anna Power Station Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1, March 2018, NRC ADAMS Accession No. ML18093A445.
[29] PB-PRA-20.005, Vol. 1, Rev. 2, Peach Bottom Seismic Probabilistic Risk Assessment, Fragility Modeling Notebook, August 2018.
[30] Sequoyah Nuclear Plant Seismic Probabilistic Risk Assessment in Response to 50.54(f)
Letter with Regard to NTTF 2.1, October 2019, NRC ADAMS Accession No. ML19291A003.
[31] Virgil C. Summer Nuclear Station Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1, September 2018, NRC ADAMS Accession No. ML18271A109.
E4-71
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
[32] Vogtle Electric Generating Plants 1 and 2 Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1, March 2017, NRC ADAMS Accession No. ML17088A130.
[33] Watts Bar Nuclear Plant Seismic Probabilistic Risk Assessment in Response to 50.54(f)
Letter with Regard to NTTF 2.1, June 2017, NRC ADAMS Accession No. ML17181A485.
[34] E-mail from R. Drsek (FENOC) to V. Andersen (Jensen Hughes, Exelon RM), "RE: BV failure to scram Am>2g?," 2/7/2020 [e-mail clarifies that failure to scram fragility used in Beaver Valley SPRA NTTF 2.1 submittal was set to the fragility screening level and equals Am=1.52g, Br=0.24 and Bu=0.32].
[35] CL-PRA-015, Revision 4, Clinton Power Station Probabilistic Risk Assessment, Detailed Level 2 PRA, February 2020.
[36] CL-PRA-013, Revision 6, Clinton Power Station Probabilistic Risk Assessment, Summary Notebook, February 2020.
[37] CL P&ID M05-1037 Sheet 0001, Rev. W, Fuel Pool Cooling and Clean-up.
[38] CL P&ID M05-1037 Sheet 0002, Rev. AE, Fuel Pool Cooling and Clean-up.
[39] NUREG/CR-4461, "Tornado Climatology of the Contiguous United States," Revision 2, February 2007.
[40] EPRI 3002003107, "High Wind Risk Assessment Guidelines," 2015.
[41] Exelon Nuclear, "Tornado Missile Hazard Analysis for Clinton Power Station (EC 366599, Rev. 0)," IP-S-0246, October 2007.
[42] US NRC letter from L. Rubenstein to F. Miraglia, "Safety Evaluation Report - Electric Power Research Institute (EPRI) Topical Report Concerning Tornado Missile Probabilistic Risk Assessment (PRA) Methodology," October 26, 1983.
[43] Clinton Power Station Unit 01, EC 366599 (type DCP/MOD).
[44] NRC Regulatory Issue Summary (RIS) 2015-06, "Tornado Missile Protection," June 10, 2015.
[45] NRC Enforcement Guidance Memorandum (EGM) 15-002, "Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance," June 10, 2015.
E4-72
 
ENCLOSURE 4 Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models
[46] Clinton Power Station Unit 01, EC 620942, "Tornado Generated Wind/Missile Vulneratiblity Evaluation - Phase II," December 2017.
[47] Clinton Power Station Flood Hazard Reevaluation Report, Enclosure 1, March 12, 2014 (ML14079A420).
[48] NRC Letter, Clinton Power Station, Unit No. 1 - Interim Staff Response to Reevaluated Flood Hazards submitted in Response to 10 CFR 50.54(f) Information Request - Flood-Causing Mechanism Reevaluation (TAC No. MF3654), September 3, 2015 (ML15212A010).
[49] NRC Letter, Clinton Power Station, Unit No. 1 - Correction to Staff Assessment of Response to Request for Information Pursuant to 10 CFR 50.54(f) - Flood-Causing Mechanisms Reevaluation (TAC NO. MF3654), November 18, 2015.
[50] Clinton Power Station Letter, Mitigating Strategies Flood Hazard Assessment (MSFHA)
Submittal, March 24, 2016 (ML16084A859).
[51] Risk Assessment of Operational Events, Volume 2 - External Events - Internal Fires -
Internal Flooding - Seismic - Other External Events - Frequencies of Seismically-Induced LOOP Events (RASP Handbook), Revision 1.02, US Nuclear Regulatory Commission, November 2017.
[52] CL-PRA-001, Revision 4, Clinton Power Station Probabilistic Risk Assessment, Initiating Events Notebook, February 2020.
[53] ER-AA-340, "GL 89-13 Program Implementing Procedure," Revision 9.
[54] IP Nuclear Station Engineering Department Calculation MAD 91-073 Revision 3, "Control Room Habitability Study - Anhydrous Ammonia," December 1994.
[55] 2013 Clinton Power Station Hazardous Chemical Survey, VC-94, Revision 0.
E4-73
 
ENCLOSURE 5 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF)
 
ENCLOSURE 5 Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF)
: 1. Introduction Section 4.0, Item 6 of the Nuclear Regulatory Commission's (NRC) Final Safety Evaluation (Reference 1) for NEI 06-09, Revision 0A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," (Reference 2) requires that the license amendment request (LAR) provide the plant-specific total CDF and LERF to confirm applicability of the limits of Regulatory Guide (RG) 1.174, Revision 1 (Reference 3). (Note that RG 1.174, Revision 2 [Reference 4], issued by the NRC in May 2011, did not revise these limits.)
The purpose of this enclosure is to demonstrate that the Clinton Power Station (CPS) total Core Damage Frequency (CDF) and total Large Early Release Frequency (LERF) are below the guidelines established in RG 1.174. RG 1.174 does not establish firm limits for total CDF and LERF, but it recommends that risk-informed applications be implemented only when the total plant risk is no more than about 1E-4/year for CDF and 1E-5/year for LERF. Demonstrating that these limits are met confirms that the risk metrics of NEI 06-09 can be applied to the CPS Risk-Informed Completion Time (RICT) Program.
: 2. Technical Approach Table E5-1 lists the CPS CDF and LERF point estimate values that resulted from a quantification of the baseline internal events (including internal flooding) and fire Probabilistic Risk Assessment (PRA) models (References 5 and 6, respectively). This table also includes an estimate of the seismic contribution to CDF and LERF based on the methodology detailed in CL-MISC-025, Revision 1, "Seismic CDF and LERF Estimates for the Clinton TSTF-505 (RICT)
Program," (Reference 7). Other external hazards are below accepted screening criteria and therefore do not contribute significantly to the totals.
Table E5-1 Total Baseline CDF/LERF Clinton Unit 1 Baseline CDF                    Clinton Unit 1 Baseline LERF Source            Contribution                  Source              Contribution Internal Events PRA          3.3E-06            Internal Events PRA          1.7E-07 Fire PRA          7.8E-05                        Fire PRA          5.3E-06 Seismic          6.4E-06                        Seismic          1.6E-06 No significant                                  No significant Other External Events                            Other External Events contribution                                    contribution Total Unit 1 CDF          8.8E-05              Total Unit 1 LERF          7.1E-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, CPS TSTF-505 implementation is consistent with NEI 06-09 guidance.
E5-1
 
ENCLOSURE 5 Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF)
: 3. References
: 1. Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,'"
dated May 17, 2007 (ADAMS Accession No. ML071200238).
: 2. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"
Revision 0, dated October 12, 2012 (ADAMS Accession No. ML12286A322).
: 3. Regulatory Guide 1.174, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, November 2002.
: 4. Regulatory Guide 1.174, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, May 2011 (Accession No. ML10091006).
: 5. CL-PRA-014, Revision 9, Clinton Power Station Probabilistic Risk Assessment Quantification Notebook, February 2020.
: 6. CL-PRA-021.11, Revision 2, Clinton Power Station Fire PRA Summary & Quantification Notebook, February 2020
: 7. CL-MISC-025, Revision 1, Seismic CDF and LERF Estimates for the Clinton TSTF-505 (RICT) Program, March 2020 E5-2
 
ENCLOSURE 6 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Justification of Application of At-Power PRA Models to Shutdown Modes This enclosure is not applicable to the Clinton Power Station submittal. Exelon Generation Company, LLC is proposing to apply the Risk-Informed Completion Time Program only in Modes 1 and 2 and not in the shutdown Modes.
 
ENCLOSURE 7 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 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
 
ENCLOSURE 7 PRA Model Update Process
: 1. Introduction Section 4.0, Item 8 of the Nuclear Regulatory Commission's (NRC) Final Safety Evaluation (Reference 1) for NEI 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," (Reference 2) requires that the license amendment request (LAR) provide a discussion of the licensee's programs and procedures which assure the PRA models which support the RMTS are maintained consistent with the as-built/as-operated plant.
This enclosure describes the administrative controls and procedural processes applicable to the configuration control of PRA models used to support the Risk-Informed Completion Time (RICT)
Program, which will be in place to ensure that these models reflect the as-built/as-operated plant.
Plant changes, including physical modifications and procedure revisions, will be identified and reviewed prior to implementation to determine if they could impact the PRA models per ER- AA-600-1015, FPIE [Full Power Internal Events] PRA Model Update (Reference 3), and ER-AA-600-1061, Fire PRA Model Update and Control (Reference 4). The configuration control program will ensure these plant changes are incorporated into the PRA models as appropriate. The process will include discovered conditions associated with the PRA models, which will be addressed by the applicable site Corrective Action Program.
Should a plant change or a discovered condition be identified that has a significant impact to the RICT Program calculations as defined by the above procedures, an unscheduled update of the PRA model will be implemented. Otherwise, the PRA model change is incorporated into a subsequent periodic model update. Such pending changes are considered when evaluating other changes until they are fully implemented into the PRA models. Periodic updates are typically performed every two refueling cycles.
: 2. PRA Model Update Process Internal Event, Internal Flood, and Fire PRA Model Maintenance and Update The Fleet risk management process ensures that the applicable PRA models used for the RICT Program reflect the as-built/as-operated plant for Clinton Unit 1. The PRA configuration control process delineates the responsibilities and guidelines for updating the full power internal events, internal flood, and fire PRA models, and includes both periodic and unscheduled PRA model updates.
The process includes provisions for monitoring potential impact areas affecting the technical elements of the PRA models (e.g., due to plant changes, plant/industry operational experience, or errors or limitations identified in the model), assessing the individual and cumulative risk impact of unincorporated changes, and controlling the model and necessary computer files, including those associated with the real time risk model.
Changes that are considered an upgrade per the ASME/ANS PRA standard receive a peer review focused on those aspects of the PRA model that represent the upgrade.
E7-1
 
ENCLOSURE 7 PRA Model Update Process Review of Plant Changes for Incorporation into the PRA Model
: 1. Plant changes or discovered conditions are reviewed for potential impact to the PRA models, including the real time risk model and the subsequent risk calculations which support the RICT Program (NEI 06-09, Section 2.3.4, Items 7.2 and 7.3, and 2.3.5, Items 9.2 and 9.3).
: 2. Plant changes that meet the criteria defined in References 3 and 4 (including consideration of the cumulative impact of other pending changes) will be incorporated in the applicable PRA model(s), consistent with the NEI 06-09 guidance. Otherwise, the change is assigned a priority and is incorporated at a subsequent periodic update consistent with procedural requirements. (NEI 06-09, Section 2.3.5, Item 9.2)
: 3. PRA updates for plant changes are performed at least once every two refueling cycles, consistent with the guidance of NEI 06-09 (NEI 06-09, Section 2.3.4, Item 7.1, and 2.3.5, Item 9.1).
: 4. If a PRA model change is required for the real time risk model, but cannot be immediately implemented for a significant plant change or discovered condition, either:
: a. Interim analyses to address the expected risk impact of the change will be performed.
In such a case, these interim analyses become part of the RICT Program calculation process until the plant changes are incorporated into the PRA model during the next update. The use of such bounding analyses is consistent with the guidance of NEI 06-09.
OR
: b. Appropriate administrative restrictions on the use of the RICT Program for extended Completion Times are put in place until the model changes are completed, consistent with the guidance of NEI 06-09.
These actions satisfy NEI 06-09, Section 2.3.5, Item 9.3.
: 3. References
: 1. Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,'"
dated May 17, 2007 (ADAMS Accession No. ML071200238).
: 2. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"
Revision 0-A, dated October 12, 2012 (ADAMS Accession No. ML12286A322).
: 3. ER-AA-600-1015, "FPIE PRA Model Update."
: 4. ER-AA-600-1061, "Fire PRA Model Update and Control."
E7-2
 
ENCLOSURE 8 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Attributes of the Real Time Risk Model
 
ENCLOSURE 8 Attributes of the Real Time Risk Model
: 1. Introduction Section 4.0, Item 9 of the Nuclear Regulatory Commission's (NRC) Final Safety Evaluation (Reference 1) for NEI 06-09, Revision 0-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines," (Reference 2) requires that the license amendment request (LAR) provide a description of PRA models and tools used to support the RMTS. This includes identification of how the baseline probabilistic risk assessment (PRA) model is modified for use in the Configuration Risk Management Program (CRMP) tools, quality requirements applied to the PRA models and CRMP tools, consistency of calculated results from the PRA model and the CRMP tools, and training and qualification programs applicable to personnel responsible for development and use of the CRMP tools. NEI 06-09, Revision 0-A, uses the term CRMP for the program controlling the use of RMTS. This term is also used to designate the program implementing 10 CFR 50.65(a)(4) and the monitoring program for other risk informed LARs. To avoid confusion the term RICT program is used to indicate the program required by NEI 06-09, Revision 0-A, in lieu of the term CRMP. This item should also confirm that the RICT program tools can be readily applied for each Technical Specification (TS) Limiting Condition for Operation (LCO) within the scope of the plant-specific submittal.
This enclosure describes the necessary changes to the peer-reviewed baseline PRA models for use in the Real Time Risk (RTR) tool to support the Risk-Informed Completion Time (RICT)
Program. The process employed to adapt the baseline models is demonstrated:
a) to preserve the Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) quantitative results; b) to maintain the quality of the peer-reviewed PRA models; and c) to correctly accommodate changes in risk due to configuration-specific considerations.
Quality controls and training programs applicable for the RICT Program are also discussed in this enclosure.
: 2. Translation of Baseline PRA Model for Use in Configuration Risk The baseline PRA models for internal events, including internal flood and internal fire, are the peer-reviewed models. These models are updated when necessary to incorporate plant changes to reflect the as-built/as-operated plant. The internal flood model is integrated into the internal events model. These models will be used in the RICT Program. The models may be optimized for quantification speed but are verified to provide the same result as the baseline models in accordance with approved procedures.
The RTR tool will be used to facilitate all configuration-specific risk calculations and support the RICT Program implementation. The PRA Models utilize system initiator event fault trees so equipment unavailabilities are captured explicitly in these system initiator fault trees.
Therefore, no adjustment to initiating event frequencies in required within the RTR tool.
E8-1
 
ENCLOSURE 8 Attributes of the Real Time Risk Model 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 configuration.
* Mutually exclusive combinations, including normally disallowed maintenance combinations, are adjusted to allow accurate analysis of the configuration.
* For systems where some trains or components are in service and some in standby or there are seasonal dependencies, the RTR tool addresses the actual configuration of the plant.
* RHR recovery terms set to one/true to remove credit.
* There are no changes in success criteria based on the time in the core operating cycle.
The configuration risk software is designed to quantify the configuration for both internal events, including internal flooding and fire, and includes the seismic risk contribution when calculating the Risk-Management Action Time (RMAT) and RICT. Full quantifications will be used for each configuration. Pre-solved cutsets will be limited to results for specific configurations. For configurations without pre-solved cutsets the model will be quantified to produce cutsets for the previously unanalyzed configuration. If there are any changes in the underlying PRA, the PRA Results database in PARAGON will be updated in accordance with the RTR Update Procedure. The unique aspect of the configuration risk software for the RICT program is the quantification of fire risk and the inclusion of the seismic risk contribution. The other adjustments above are those used for the evaluation of risk under the 10CFR 50.65(a)(4) program.
The Clinton Power Station (CPS) Unit 1 PRA calculates Common Cause Basic Event (CCBE) probabilities from alpha factors and places the basic events under appropriate gates in the fault tree.
Adjustments to the Common Cause Failure (CCF) grouping or CCF probabilities are not necessary when a component is taken out-of-service for preventative maintenance:
* The component is not out-of-service for reasons subject to a potential common cause failure, and so the in-service components are not subject to increases in common cause probabilities.
* CCF relationships are retained for the remaining in-service components.
* The net failure probability for the in-service components includes the CCF contribution of the out-of-service component.
As described in Reg Guide 1.177 (Reference 6), Section A-1.3.2.2, the CCF term should be treated differently when a component is taken down for Preventive Maintenance (PM) than as described for failure of a component. For PMs, the common cause factor is changed so that the model represents the unavailability of the remaining component. In the example provided in Reg Guide 1.177 for a 2-train system, the CCF event can be set to zero for PMs. This is done so that the model represents the unavailability of the remaining component, and not the common cause multiplier. The CPS approach is conservative in that for a 2-train system, the CCF event is retained for the component removed from service. Likewise, for systems with E8-2
 
ENCLOSURE 8 Attributes of the Real Time Risk Model three or more trains, the CCF events that are related to the out-of-service component are retained.
The Vogtle RICT Safety Evaluation (Reference 5) describes the Vogtle approach for modeling common cause events with planned inoperability: "For planned inoperability, the licensee sets the appropriate independent failure to 'true' and makes no other changes while calculating a RICT." The CPS approach is the same as this Vogtle approach.
It is recognized that other modifications could be made to CCF factors for planned maintenance, particularly for common cause groups of three or more components. For example, in the Safety Evaluation (SE) in the Vogtle RICT Amendment (Reference 5), the NRC identifies a possible planned maintenance CCF modification to "modify all the remaining basic event probabilities to reflect the reduced number of redundant components."
Like Vogtle, the CPS CCF approach is a straightforward simplification that has inherent uncertainties. In the context of modifying CCF basic events for PMs, the Vogtle SE states the following:
        "The NRC staff also notes that common cause failure probability estimates are very uncertain and retaining precision in calculations using these probabilities will not necessarily improve the accuracy of the results. Therefore, the NRC staff concludes that the licensee's method is acceptable because it does not systematically and purposefully produce non-conservative results and because the calculations reasonably include common cause failures consistent with the accuracy of the estimates." (Reference 5)
The CPS approach for CCF during PMs is the same as the Vogtle approach; therefore, the CPS CCF approach is acceptable for RICT calculations and adjusting the common cause grouping is not necessary for PMs. However, if a numeric adjustment is performed, the RICT calculation shall be adjusted to numerically account for the increased possibility of CCF in accordance with RG 1.177, as specified in Section A-1.3.2.1 of Appendix A of the RG.
For emergent conditions where the extent of condition is not completed prior to entering into the Risk Management Action Times or the extent of condition cannot rule out the potential for common cause failure, common cause RMAs are expected to be implemented to mitigate common cause failure potential and impact, in accordance with Exelon procedures. This is in line with the guidance of NEI 06-09 and precludes the need to adjust CCF probabilities.
However, if a numeric adjustment is performed, the RICT calculation shall be adjusted to numerically account for the increased possibility of CCF in accordance with RG 1.177, as specified in Section A-1.3.2.1 of Appendix A of the RG.
: 3. Quality Requirements and Consistency of PRA Model and Configuration Risk Tools The approach for establishing and maintaining the quality of the PRA models, including the configuration risk model, includes both a PRA maintenance and update process (described in ), and the use of self-assessments and independent peer reviews (described in ).
E8-3
 
ENCLOSURE 8 Attributes of the Real Time Risk Model The information provided in Enclosure 2 demonstrates that the site's internal event, internal flood, and internal fire PRA models reasonably conform to the associated industry standards endorsed by Regulatory Guide 1.200 (Reference 3). This information provides a robust basis for concluding that the PRA models are of sufficient quality for use in risk-informed licensing actions.
For maintenance of an existing configuration risk model, changes made to the baseline PRA model in translation to the configuration risk model will be controlled and documented. Every PRA Model of Record (MOR) Update results in an update to the RTR model in accordance with the FPIE and Fire PRA Update procedures. An acceptance test is performed after every configuration risk model update. This testing also verifies correct mapping of plant components to the basic events in the configuration risk model. The RTR model documentation includes changes made to the MOR model files to work with the RTR model software (e.g., quantification settings) along with verification that results are consistent between the RTR and PRA zero maintenance results. In addition, the RTR update for the MOR includes quantifying the RTR model for representative maintenance configurations and examining the results for appropriateness. These actions are procedurally controlled.
: 4. Training and Qualification The PRA staff is responsible for development and maintenance of the configuration risk model.
Operations and Work Control staff will use the configuration risk tool under the RICT Program.
PRA Staff and Operations are trained in accordance with a program using National Academy for Nuclear Training (ACAD) documents, which is also accredited by Institute of Nuclear Power Operations (INPO).
: 5. Application of the Configuration Risk Tool to the RICT Program Scope The PARAGON software will be used to facilitate all configuration-specific risk calculations and support the RICT Program implementation. This program is specifically designed to support implementation of RMTS. PARAGON will permit the user to evaluate all plant configurations using appropriate mapping of equipment to PRA basic events. The equipment in the scope of the RICT program will be able to be evaluated in the appropriate PRA models. The RICT program will meet RG 1.174 (Reference 4) and Exelon software quality assurance requirements.
: 6. References
: 1. Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI),"Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,'"
dated May 17, 2007 (ADAMS Accession No. ML071200238).
: 2. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"
Revision 0-A, dated October 12, 2012 (ADAMS Accession No. ML12286A322).
: 3. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, March 2009.
E8-4
 
ENCLOSURE 8 Attributes of the Real Time Risk Model
: 4. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, May 2011.
: 5. Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments Regarding Implementation of Topical Report Nuclear Energy Institute NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specification (RMTS)
Guidelines," Revision 0-A (CAC NOS. ME9555 and ME9556), ML15127a669.
: 6. Nuclear Regulatory Commission, Regulatory Guide 1.177, May 2011, Revision 1.
E8-5
 
ENCLOSURE 9 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 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
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty
: 1. Introduction The purpose of this enclosure is to disposition the impact of Probabilistic Risk Assessment (PRA) modeling epistemic uncertainty for the Risk Informed Completion Time (RICT) Program.
Topical Report NEI 06-09-A (Reference 1), Section 2.3.4, item 10 requires an evaluation to determine insights that will be used to develop risk management actions (RMAs) to address these uncertainties. The baseline Internal Events PRA and Fire PRA (FPRA) models document assumptions and sources of uncertainty and these were reviewed during the model peer reviews. The approach taken is, therefore, to review these documents to identify the items which may be directly relevant to the RICT Program calculations, to perform sensitivity analyses where appropriate, to discuss the results and to provide dispositions for the RICT Program.
The epistemic uncertainty analysis approach described below applies to the Internal Events PRA and any epistemic uncertainty impacts that are unique to FPRA are also addressed. In addition, Topical Report NEI 06-09-A requires that the uncertainty be addressed in RICT Program Configuration Risk Management Program (CRMP), otherwise referred to as the Real-Time Risk (RTR), tools by consideration of the translation from the PRA model to the RTR model. The RTR model, also referred to as the PARAGON model, discussed in Enclosure 8 includes internal events, flooding events and fire events. The model translation uncertainties evaluation and impact assessment are limited to new uncertainties that could be introduced by application of the RTR tool during RICT Program calculations.
: 2. Assessment of Internal Events PRA Epistemic Uncertainty Impacts In order to identify key sources of uncertainty for the RICT Program application, the Internal Events PRA model uncertainties were evaluated using the guidance in NUREG-1855 (Reference 2) and Electric Power Research Institute (EPRI) 1016737 (Reference 3). As described in NUREG-1855, sources of uncertainty include "parametric" uncertainties, "modeling" uncertainties, and "completeness" (or scope and level of detail) uncertainties.
Parametric uncertainty was addressed as part of the Clinton Power Station (CPS) PRA model quantification. The parametric uncertainty evaluation for the Internal Events PRA model is documented in the Summary Notebook (Reference 4).
Modeling uncertainties are considered in both the base PRA and in specific risk-informed applications. Assumptions are made during the PRA development to address modeling uncertainties because there is not a single definitive approach. Plant-specific assumptions made for each of the CPS Internal Events PRA technical elements are noted in the individual notebooks and summarized in Reference 4. The Internal Events PRA model uncertainties evaluation considers the modeling uncertainties for the base PRA by identifying assumptions, determining if those assumptions are related to a source of modeling uncertainty and characterizing that uncertainty, as necessary. EPRI compiled a listing of generic sources of modeling uncertainty to be considered for each PRA technical element (Reference 3), and the evaluation performed for CPS considered each of the generic sources of modeling uncertainty as well as the plant-specific sources.
E9-1
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Completeness uncertainty addresses scope and level of detail. Uncertainties associated with scope and level of detail are documented in the PRA but are only considered for their impact on a specific application (Reference 4). No specific issues of PRA completeness have been identified relative to the TSTF-505 application, based on the results of the Internal Events PRA peer reviews.
Additionally, an evaluation of Level 2 internal events PRA model uncertainty was performed, based on the guidance in NUREG-1855 (Reference 2) and Electric Power Research Institute (EPRI) 1026511 (Reference 5). The potential sources of model uncertainty in the CPS PRA model were evaluated for the 32 Level 2 PRA topics outlined in EPRI 1026511.
A detailed review of the generic and plant-specific sources of internal events model uncertainties is discussed in CL-MISC-033 (Reference 7) and are therefore not repeated in this enclosure. The purpose of this enclosure is to summarize the key sources of uncertainty that could potentially impact the RICT calculations.
Based on following the methodology in EPRI 1016737, as supplemented by EPRI 1026511, the impact of key sources of uncertainty in the Internal Events PRA model on the RICT application is summarized in Table E9-1. The key sources of uncertainty identified in Table E9-1 do not present a significant impact on the CPS RICT calculations and therefore, the Internal Events PRA model can produce accurate RICT calculations. Note that RMAs will be developed when appropriate using insights from the PRA model results specific to the configuration.
E9-2
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-1 ASSESSMENT OF INTERNAL EVENTS PRA EPISTEMIC UNCERTAINTY Source of Uncertainty and Assumptions                      RICT Program Impact                          Model Sensitivity and Disposition ECCS Survivability Post Containment Venting Failure ECCS survivability post-        Although the treatment is realistic, there  For this source of model uncertainty, sensitivity analyses were containment venting is          is the potential for a non-conservative      performed for a select group of technical specifications within scope treated probabilistically.      bias given the unknown                      of the RICT application.
phenomenological events that could be associated with containment venting          For this sensitivity case, ECCS failure probability due to steam (e.g., hydrogen buildup in the Auxiliary    binding was escalated by a factor of ten (10) (i.e., from 0.01 to 0.1)
Building, harsh events due to steam          and catastrophic containment failure probability was escalated by a release, and other unknown                  factor of two (2) (i.e., from 0.2 to 0.4).
consequences).
Although some of the RICT estimates change as a result of this Therefore, this assumption was              sensitivity, the bounding sensitivity analysis assumes the upper identified as a candidate source of          bound value for ECCS equipment failure due to steam binding, which model uncertainty and was further            is not a realistic assumption and use of this bounding assumption evaluated with various sensitivity          would result in overly-conservative RICT estimates.
analyses.
Therefore, the uncertainty associated with this model uncertainty is negligible within the RICT application.
Core Melt Arrest Prior to Vessel Failure Core melt arrest prior to        Core melt arrest prior to vessel failure    For this source of model uncertainty, sensitivity analyses were vessel failure may be credited  may not be guaranteed with LP injection      performed for a select group of technical specifications within scope to some degree with LP          recovered after core damage, but prior      of the RICT application.
injection recovered after core  to vessel failure.
damage, but prior to vessel                                                  For this sensitivity case, LP ECCS is assumed to be inadequate and failure. However, credit for    Therefore, this assumption was              vessel failure is assumed to occur (i.e., failure probability of 1.0).
the in-vessel arrest is limited  identified as a candidate source of to only a short amount of in-    model uncertainty and was further vessel core melt progression.                                                Due to the negligible impact demonstrated by the sensitivity analyses, evaluated with various sensitivity          the uncertainty associated with this model uncertainty is negligible analyses.                                    within the RICT application.
E9-3
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-1 ASSESSMENT OF INTERNAL EVENTS PRA EPISTEMIC UNCERTAINTY Source of Uncertainty and Assumptions                    RICT Program Impact                            Model Sensitivity and Disposition Vapor Suppression Capabilities at Vessel Failure Ex-vessel core melt            Ex-vessel core melt progression                For this source of model uncertainty, sensitivity analyses were progression overwhelms        overwhelms vapor suppression                  performed for a select group of technical specifications within scope vapor suppression noted as    capabilities is identified as a potential      of the RICT application.
extremely unlikely for low    candidate source of model uncertainty.
pressure RPV failures modes                                                  For this sensitivity case, the recommended upper bound values from and very unlikely for high    It is noted that NUREG/CR-6595 would          NUREG/CR-6595 (Reference 9) for Mark III Containments were used pressure failure modes based  indicate that upper bound values of 0.2        as an alternate hypothesis (i.e., sensitivity analysis uses upper bound on reference to generic        for both low & high pressure scenarios        values of 0.2 for low pressure and high pressure scenarios).
studies and identification of  (Mark III specific values) may need to plant-specific features.      be explored as an alternate hypothesis.        Due to the negligible impact demonstrated by the sensitivity analyses, the uncertainty associated with this model uncertainty is negligible Therefore, this assumption was                within the RICT application.
identified as a candidate source of model uncertainty and was further evaluated with various sensitivity analyses.
Hydrogen Combustion The Mark III containment is    The hydrogen combustion logic in the          For this source of model uncertainty, sensitivity analyses were not inerted. Hydrogen          Level 2 Containment Event Trees                performed for a select group of technical specifications within scope igniters are provided for      (CETs) is considered realistic, however,      of the RICT application.
controlled burn of hydrogen    this node of the CETs is critical in the buildup in containment.        estimation of accident sequence                For this sensitivity case, the hydrogen igniter failure probability was Severe accident progression    frequencies with higher radionuclide          escalated by a factor of 100 (i.e., from 1E-4 to 1E-2).
is modeled to lead to          release fractions.
hydrogen combustion which fails containment with                                                        Due to the negligible impact demonstrated by the sensitivity analyses, Therefore, this assumption was                the uncertainty associated with this model uncertainty is negligible operation of the hydrogen      identified as a candidate source of igniters.                                                                    within the RICT application.
model uncertainty and was further evaluated with various sensitivity analyses.
E9-4
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-1 ASSESSMENT OF INTERNAL EVENTS PRA EPISTEMIC UNCERTAINTY Source of Uncertainty and Assumptions                    RICT Program Impact                          Model Sensitivity and Disposition Digital Feedwater Controls There are model uncertainties  Digital feedwater control failure            For this source of model uncertainty, sensitivity analyses were associated with modeling        probabilities are treated probabilistically. performed for a select group of technical specifications within scope digital systems, such as those                                              of the RICT application.
related to determining the      Therefore, this assumption was failure modes of these          identified as a candidate source of          For this sensitivity case, the digital feedwater control failure systems and components.        model uncertainty and was further            probability was escalated by a factor of 100 (i.e., from 2E-3 to 2E-1).
evaluated with various sensitivity analyses.                                    Due to the negligible impact demonstrated by the sensitivity analyses, the uncertainty associated with this model uncertainty is negligible within the RICT application.
CRD Credit Post-Containment Failure Adverse Auxiliary Building                                                  For this source of model uncertainty, sensitivity analyses were (AB) conditions could arise as                                              performed for a select group of technical specifications within scope a result of containment                                                      of the RICT application.
failure. However, the CRD pumps are in the Turbine                                                    For this sensitivity case, the failure probability associated with CRD Building. Therefore, only                                                    credit post-containment failure was escalated by a factor of 100 (i.e.,
failure modes that would                                                    from ~2E-5 to ~2E-3).
directly impact the CRD injection lines are modeled to fail CRD following                                                          Due to the negligible impact demonstrated by the sensitivity analyses, containment failure.                                                        the uncertainty associated with this model uncertainty is negligible within the RICT application.
E9-5
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-1 ASSESSMENT OF INTERNAL EVENTS PRA EPISTEMIC UNCERTAINTY Source of Uncertainty and Assumptions                      RICT Program Impact                          Model Sensitivity and Disposition FLEX Equipment Reliability There are no industry-          Given the significance of loss of offsite    For this source of model uncertainty, sensitivity analyses were approved data sources for        power in the model, the RICT                performed for a select group of technical specifications within scope FLEX equipment reliability.      calculations may be sensitive to FLEX        of the RICT application.
equipment failure rates.
The equipment failure rate                                                    For this sensitivity case, the FLEX equipment failure probabilities data from equivalent non-        Therefore, this assumption was              were escalated by a factor of five (5), which equated to a factor of ten FLEX systems is used as a        identified as a candidate source of          (10) when compared to the original non-FLEX equipment failure rates surrogate for the FLEX          model uncertainty and was further            used for non-FLEX equipment.
equipment modeled in the        evaluated with various sensitivity PRA (until industry approved    analyses.                                    Due to the small impact demonstrated by the sensitivity analyses, the FLEX data is developed).                                                      uncertainty associated with this model uncertainty is negligible within the RICT application.
Water Hammer Pipe Rupture Water hammer is a potential      The water hammer consequences (i.e.,        For this source of model uncertainty, sensitivity analyses were failure mode for ECCS that      flow blockage, leakage, and rupture)        performed for a select group of technical specifications within scope may result in a large flood in  are treated probabilistically.              of the RICT application.
the Auxiliary Building (AB) basement.                        Therefore, this assumption was              For this sensitivity case, the ECCS pipe failure probabilities due to identified as a candidate source of          blockage, leakage, and rupture by water hammers were escalated by ECCS draindown scenarios        model uncertainty and was further            a factor of five (5) (i.e., combined failure probability increased to ~0.5 are included in the PRA          evaluated with various sensitivity          from ~0.1).
model. Subsequent starting /    analyses.
restarting of these systems                                                  Due to the small impact demonstrated by the sensitivity analyses, the (with or without starting water                                              uncertainty associated with this model uncertainty is negligible within leg pumps) can cause a                                                        the RICT application.
water hammer and flooding of the AB basement.
E9-6
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-1 ASSESSMENT OF INTERNAL EVENTS PRA EPISTEMIC UNCERTAINTY Source of Uncertainty and Assumptions                  RICT Program Impact                          Model Sensitivity and Disposition Containment Integrity Following Vessel Rupture There is model uncertainty    Containment integrity following vessel      For this source of model uncertainty, sensitivity analyses were regarding the subsequent      rupture is treated probabilistically.        performed for a select group of technical specifications within scope treatment that increases the                                              of the RICT application.
likelihood of LERF for this  Therefore, this assumption was extremely rare event.        identified as a candidate source of          For this sensitivity case, containment failure probability was escalated model uncertainty and was further            by a factor of 100 (i.e., from 1E-5 to 1E-3).
A portion of the vessel      evaluated with various sensitivity rupture sequences are        analyses.                                    Due to the negligible impact demonstrated by the sensitivity analyses, assumed to result in                                                      the uncertainty associated with this model uncertainty is negligible concurrent containment                                                    within the RICT application.
failure coincident with the vessel rupture.
E9-7
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty
: 3. Assessment of Translation (RTR Model) Uncertainty Impacts Incorporation of the baseline PRA models into the RTR model used for RICT Program calculations may introduce new sources of model uncertainty. Table E9-2 provides a description of the relevant model changes and dispositions of whether any of the changes made represent possible new sources of model uncertainty that must be addressed. Refer to Enclosure 8 for additional discussion on the RTR model.
Table E9-2 ASSESSMENT OF TRANSLATION UNCERTAINTY IMPACTS RTR Model Change      Part of Model and Assumptions        Affected                  Impact on Model          Disposition PRA model logic        Fault tree logic model    The model, if            Since the restructured structure may be      structure, affecting both  restructured, will be    model will produce optimized to          internal and Fire PRAs. logically equivalent and comparable numerical increase solution                                produce results          results, this is not a source speed.                                            comparable to the        of uncertainty for the RICT baseline PRA logic      program.
model.
Incorporation of      Calculation of RICT        The addition of          Since this is a bounding seismic risk bias to  and Risk Management        bounding impacts for    approach for addressing support RICT          Action Threshold          seismic events has no    seismic risk in the RICT Program risk          (RMAT) within RTR.        impact on baseline      Program it is not a source calculations.                                    PRA or RTR model.        of translation uncertainty Impact is reflected in  and RICT Program A conservative value                              calculation of all RICTs calculations are not for the seismic delta                            and RMATs.              impacted, so no mandatory CDF is applicable.                                                        Risk Management Actions (RMAs) are required.
Set plant availability Basic Event:              Since the RTR model      This change is consistent (Reactor Critical      1--SYAVAILFAC---          evaluates specific      with RTR tool practice; Years Factor) basic                              configurations during    therefore, this change does event to 1.0.                                    at-power conditions,    not represent a source of the use of a plant      uncertainty, and RICT availability factor less program calculations are than 1.0 is not          not impacted, so no appropriate. This        mandatory RMAs are change allows the RTR    required.
model to produce appropriate results for specific at-power configurations.
RHR Recovery          Basic Events:              Setting these terms to  Not taking credit for RHR Terms                  1RHRX-REC-AT-F--          1.0 prevents any credit  recovery removes a 1RHRX-REC-UPDH--          being taken for RHR      potential source of recovery.                uncertainty.
1RHRXDHRRECLTH--
1RHRXDHRRECLTH-F E9-8
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty
: 4. Assessment of Supplementary FPRA Epistemic Uncertainty Impacts The Fire PRA (FPRA) includes various sources of uncertainty that exist because there are both inherent randomness in elements that comprise the FPRA and because the state of knowledge in these elements continues to evolve. The development of the CPS FPRA was developed using consensus modeling approaches described within NUREG/CR-6850 (Reference 10).
In order to identify key sources of uncertainty for the RICT Program application, the FPRA model uncertainties were evaluated using the guidance in NUREG-1855 (Reference 2) and Electric Power Research Institute (EPRI) 1026511 (Reference 5). As described in NUREG-1855, sources of uncertainty include "parametric" uncertainties, "modeling" uncertainties, and "completeness" (or scope and level of detail) uncertainties.
Parametric uncertainty was addressed as part of the Clinton Power Station (CPS) FPRA model quantification. The parametric uncertainty evaluation for the FPRA model is documented in the Uncertainty and Sensitivity Analysis Notebook (Reference 6).
Modeling uncertainties are considered in both the base FPRA and in specific risk-informed applications. Assumptions are made during the FPRA development to address modeling uncertainties because there is not a single definitive approach. Plant-specific assumptions made for each of the CPS FPRA technical elements are noted in the individual notebooks and summarized in Reference 6. The FPRA model uncertainties evaluation considers the modeling uncertainties for the base FPRA by identifying assumptions, determining if those assumptions are related to a source of modeling uncertainty and characterizing that uncertainty, as necessary. EPRI compiled a listing of generic sources of modeling uncertainty to be considered for each FPRA technical element (Reference 5), and the evaluation performed for CPS considered each of the generic sources of modeling uncertainty as well as the plant-specific sources.
Completeness uncertainty addresses scope and level of detail. Uncertainties associated with scope and level of detail are documented in the FPRA but are only considered for their impact on a specific application (Reference 6). No specific issues of FPRA completeness have been identified relative to the TSTF-505 application, based on the results of the FPRA peer reviews.
The plant-specific assumptions in the CPS FPRA and the 71 generic sources of uncertainty identified in EPRI 1026511 were evaluated for their potential impact on the RICT application.
This guideline organizes the uncertainties in Topic Areas like those outlined in NUREG/CR-6850 and was used to evaluate the baseline FPRA epistemic uncertainty and evaluate the impact of this uncertainty on RICT Program calculations.
A detailed review of the generic and plant-specific sources of internal fire model uncertainties are discussed in CL-MISC-033 (Reference 7) and are therefore not repeated in this enclosure.
The purpose of this enclosure is to summarize the key sources of uncertainty that could potentially impact the RICT calculations.
Table E9-3 summarizes the review for key sources of uncertainty in the FPRA model for the RICT application (organized by NUREG/CR-6850 tasks).
E9-9
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty As noted above, the CPS FPRA was developed using consensus methods outlined in NUREG/CR-6850 and interpretations of technical approaches as required by NRC. FPRA methods were based on NUREG/CR-6850, other more recent NUREGs, (e.g., NUREG-7150 (Reference 11)), and published "frequently asked questions" (FAQs).
The key sources of uncertainty identified in Table E9-3 do not present a significant impact on the CPS RICT calculations and therefore, the FPRA model can produce accurate RICT calculations. Note that RMAs will be developed when appropriate using insights from the PRA model results specific to the configuration.
E9-10
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-3 ASSESSMENT OF FIRE PRA EPISTEMIC UNCERTAINTY Task
#  Description  Sources of Uncertainty                            Disposition for RICT Application 1  Analysis    This task establishes the overall spatial scope  Based on a review of the assumptions and potential sources of sources boundary and of the analysis and provides a framework for      of uncertainly associated with this element, it is concluded that the partitioning organizing the data for the analysis. The        methodology for the Analysis Boundary and Partitioning task does not partitioning features credited are required to    introduce any epistemic uncertainties that would affect the RICT satisfy established industry standards.          calculation.
Therefore, RICT Program calculations are not impacted, and no RMAs are required to address this item.
2  Component    This task involves the selection of              The uncertainty associated with this task is related to the identification Selection    components to be treated in the analysis in      of all components that should be credited/linked in the FPRA. This the context of initiating events and mitigation. source of uncertainty is reduced as a result of multiple overlapping The potential sources of uncertainty include      tasks including the MSO expert panel, reviews of FPIE screened those inherent in the internal events PRA        initiating events, screened containment penetrations, and screened model as that model provides the foundation      ISLOCA scenarios. Additional internal reviews of analysis results for the Fire PRA.                                further reduce the uncertainty associated with this task.
Based on a review of the assumptions and potential sources of uncertainty related to this element and the discussion above, it is concluded that the methodology for the Component Selection task does not introduce any epistemic uncertainties that would affect the RICT calculation.
Therefore, RICT Program calculations are not impacted, and no RMAs are required to address this item.
E9-11
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-3 ASSESSMENT OF FIRE PRA EPISTEMIC UNCERTAINTY Task
#  Description Sources of Uncertainty                          Disposition for RICT Application 3  Cable      The selection of cables to be considered in      Additionally, as part of the Fire PRA, some components were Selection  the analysis is identified using industry        conservatively assumed to be failed based on lack of cable data.
guidance documents. The overall process is      Components in this category are referred to as Unknown Location essentially the same as that used to perform    (UNL) components because specific cables were not identified for the the analyses to demonstrate compliance with      components. Based on recent Fire PRA updates, the UNL components 10 CFR 50.48.                                    are mostly limited to Balance of Plant (BOP) systems.
Two sensitivity analyses were performed to measure the risk associated with the assumption that these components fail in select fire scenarios. The first sensitivity removed all UNL components from every fire scenario (estimates potential conservatisms) and the second sensitivity evaluated expanded UNL failures in every fire scenario (estimates potential non-conservatisms). The sensitivity analyses are documented in the Uncertainty and Sensitivity Analysis Notebook (Reference 6).
Based on the results of these sensitivity analyses, the UNL methodology does not introduce significant conservatisms into the base FPRA model and is assessed to be appropriate to avoid overly conservative results that mask key risk insights. Given that an informed approach was used to develop the assumed routing, the methodology employed by the FPRA is appropriate.
Based on a review of the assumptions and potential sources of uncertainty related to this element it is concluded that the methodology for the Cable Selection task does not introduce any epistemic uncertainties that would affect the RICT calculation.
Therefore, RICT Program calculations are not impacted, and no RMAs are required to address this item.
E9-12
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-3 ASSESSMENT OF FIRE PRA EPISTEMIC UNCERTAINTY Task
#  Description  Sources of Uncertainty                              Disposition for RICT Application 4  Qualitative  Qualitative screening was performed;                In the event a structure (location) which could result in a plant trip was Screening    however, some structures (locations) were          incorrectly excluded, its contribution to CDF would be small (with a eliminated from the global analysis boundary        CCDP commensurate with base risk). Such a location would have a and ignition sources deemed to have no              negligible risk contribution to the overall FPRA.
impact on the FPRA (based on industry guidance and criteria) were excluded from the      Based on a review of the assumptions and potential sources of quantification based on qualitative screening      uncertainty related to this element and the discussion above, it is criteria. The only criterion subject to            concluded that the methodology for the Qualitative Screening task does uncertainty is the potential for plant trip.        not introduce any epistemic uncertainties that would affect the RICT However, such locations would not contain          calculation.
any features (equipment or cables identified in the prior two tasks) and consequently are expected to have a low risk contribution.          Therefore, RICT Program calculations are not impacted, and no RMAs are required to address this item.
5  Fire-Induced The internal events PRA model was updated          The identified source of uncertainty could result in the over-estimation Risk Model  to add fire specific initiating event structure as  of fire risk. In general, the Fire PRA development process would have well as additional system logic. The                reviewed significant fire initiating events and performed supplemental methodology used is consistent with that            assessments to address this possible source of uncertainty.
used for the internal events PRA model development as was subjected to industry            Based on a review of the assumptions and potential sources of Peer Review.                                        uncertainty related to this element and the discussion above, it is concluded that the methodology for the Fire-Induced Risk Model task The developed model is applied in such a            does not introduce any epistemic uncertainties that would affect the fashion that all postulated fires are assumed      RICT calculation.
to generate a plant trip. This represents a source of uncertainty, as it is not necessarily    Therefore, RICT Program calculations are not impacted, and no RMAs clear that fires would result in a trip. In the    are required to address this item.
event the fire results in damage to cables and/or equipment identified in Task 2, the PRA model includes structure to translate them into the appropriate induced initiator.
E9-13
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-3 ASSESSMENT OF FIRE PRA EPISTEMIC UNCERTAINTY Task
#  Description  Sources of Uncertainty                            Disposition for RICT Application 6  Fire Ignition Fire ignition frequency is an area with            The FPRA utilized the bin frequencies from NUREG/CR-2169 Frequency    inherent uncertainty. Part of this uncertainty    (Reference 12), which represents the most current approved source for arises due to the counting and related            bin frequencies. As such, some of the inherent conservatism partitioning methodology.                          associated with bin frequencies from NUREG/CR-6850 was removed.
A parametric uncertainty analysis using the Money Carlo method is However, the resulting frequency is not            provided in the Uncertainty and Sensitivity Analysis Notebook particularly sensitive to changes in ignition      (Reference 6).
source counts. The primary source of uncertainty for this task is associated with the  Consensus approaches are employed in the model.
industry generic frequency values used for the Fire PRA. This is because there is no          Based on a review of the assumptions and potential sources of specific treatment for variability among plants    uncertainty related to this element it is concluded that the methodology along with some significant conservatism in        for the Fire Ignition Frequency task does not introduce any epistemic defining the frequencies, and their associated    uncertainties that would affect the RICT calculation.
heat release rates.
Therefore, RICT Program calculations are not impacted, and no RMAs are required to address this item.
7  Quantitative  Other than screening out potentially risk          Quantitative screening criteria was defined for the Clinton Fire PRA as Screening    significant scenarios (ignition sources), this    the CDF / LERF contribution of zero, such that all quantified fire task is not a source of uncertainty.              scenarios are retained. All of the results were retained in the cumulative CDF / LERF; therefore, no uncertainty was introduced as a result of this task.
Based on the discussion above, it is concluded that the methodology for the Quantitative Screening task does not introduce any epistemic uncertainties that would affect the RICT calculation.
Therefore, RICT Program calculations are not impacted, and no RMAs are required to address this item.
E9-14
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-3 ASSESSMENT OF FIRE PRA EPISTEMIC UNCERTAINTY Task
#  Description    Sources of Uncertainty                              Disposition for RICT Application 8  Scoping Fire    The framework of NUREG/CR-6850 includes              See Task 11 discussion.
Modeling        two tasks related to fire scenario development (Tasks 8 and 11). The discussion of uncertainty for both tasks is provided in the discussion for Task 11.
9  Detailed        The circuit analysis is performed using              Circuit analysis was performed as part of the deterministic post fire safe Circuit Failure standard electrical engineering principles.          shutdown analysis. Refinements in the application of the circuit Analysis        However, the behavior of electrical insulation      analysis results to the FPRA were performed on a case-by-case basis properties and the response of electrical            where the scenario risk quantification was large enough to warrant circuits to fire induced failures is a potential    further detailed analysis.
source of uncertainty. This uncertainty is associated with the dynamics of fire and the        Hot short probabilities and hot short duration probabilities as defined in inability to ascertain the relative timing of        NUREG-7150, Volume 2, based on actual fire test data, were used in circuit failures. The analysis methodology          the FPRA. The uncertainty (conservatism) which may remain in the assumes failures would occur in the worst            FPRA is associated with scenarios that do not contribute significantly to possible configuration, or if multiple circuits      the overall fire risk.
are involved, at whatever relative timing is required to cause a bounding worst-case outcome. This results in a skewing of the risk      Based on a review of the assumptions and potential sources of estimates such that they are over-estimated.        uncertainty related to this element and the discussion above, it is concluded that the methodology for the Detailed Circuit Failure Analysis task does not introduce any epistemic uncertainties that would affect the RICT calculation.
Therefore, RICT Program calculations are not impacted, and no RMAs are required to address this item.
E9-15
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-3 ASSESSMENT OF FIRE PRA EPISTEMIC UNCERTAINTY Task
#  Description    Sources of Uncertainty                            Disposition for RICT Application 10  Circuit Failure One of the failure modes for a circuit (cable)    The use of hot short failure probability and duration probability is based Mode            given fire induced failure is a hot short. A      on fire test data and associated consensus methodology published in Likelihood      conditional probability and a hot short          NUREG-7150, Volume 2.
Analysis        duration probability are assigned using industry guidance published in NUREG 7150,        Based on a review of the assumptions and potential sources of Volume 2. The uncertainty values specified        uncertainty related to this element and the discussion above, it is in NUREG-7150, Volume 2 are based on fire        concluded that the methodology for the Circuit Failure Mode Likelihood test data.                                        Analysis task does not introduce any epistemic uncertainties that would affect the RICT calculation.
Therefore, RICT Program calculations are not impacted, and no RMAs are required to address this item.
11  Detailed Fire  The application of fire modeling technology is    Consensus modeling approach is used for Detailed Fire Modeling and it Modeling        used in the Fire PRA to translate a fire          is concluded that the methodology for the Detailed Fire Modeling task initiating event into a set of consequences      does not introduce any epistemic uncertainties that would require (fire-induced failures). The performance of      sensitivity treatment.
the analysis requires a number of key input parameters. These input parameters include        Therefore, RICT Program calculations are not impacted, and no the heat release rate (HRR) for the fire, the    additional RMAs are required to address this item.
growth rate, the damage threshold for the targets, and response of plant staff (detection, fire control, fire suppression).
The fire modeling methodology itself is largely empirical in some respects and consequently is another source of uncertainty. For a given set of input parameters, the fire modeling results (temperatures as a function of distance from the fire) are characterized as having some distribution (aleatory uncertainty). The epistemic uncertainty arises from the selection of the input parameters (specifically the HRR and growth E9-16
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-3 ASSESSMENT OF FIRE PRA EPISTEMIC UNCERTAINTY Task
#  Description Sources of Uncertainty                          Disposition for RICT Application rate) and how the parameters are related to the fire initiating event. While industry guidance is available, that guidance is derived from laboratory tests and may not necessarily be representative of randomly occurring events.
The fire modeling results using these input parameters are used to identify a zone of influence (ZOI) for the fire and cables/equipment within that ZOI are assumed to be damaged. In general, the guidance provided for the treatment of fires is conservative and the application of that guidance retains that conservatism. The resulting risk estimates are also conservative.
E9-17
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-3 ASSESSMENT OF FIRE PRA EPISTEMIC UNCERTAINTY Task
#  Description  Sources of Uncertainty                            Disposition for RICT Application 12  Post-Fire    The human error probabilities (HEPs) used in      The HEPs include the consideration of degradation or loss of necessary Human        the Fire PRA were adjusted to consider the        cues due to fire. The fire risk importance measures indicate that the Reliability  additional challenges that may be present          results are somewhat sensitive to HRA model and parameter values.
Analysis    given a fire. The HEPs were obtained using        The FPRA model HRA is based on industry consensus modeling the EPRI HRA Calculator (HRAC) and                approaches for its HEP calculations, so this is not considered a included the consideration of degradation or      significant source of epistemic uncertainty.
loss of necessary cues due to fire. Given the methodology used, the impact of any                Assuming no credit for operator response is not realistic. However, the remaining uncertainties is expected to be          TSTF-505 procedure will require appropriate Risk Management Action small.                                            (RMA) focus on human performance for RICT entry (e.g., including an operator briefing on the significant human actions in the PRA that are pertinent to the configuration).
It is concluded that the methodology for the Post-Fire Human Reliability Analysis task does not introduce any epistemic uncertainties that would require sensitivity treatment. Therefore, RICT Program calculations are not impacted, and no additional RMAs are required to address this item.
13  Seismic-Fire Since this is a qualitative evaluation, there is  The qualitative assessment of seismic-induced fires should not be a Interactions no quantitative impact with respect to the        source of model uncertainty as it is not expected to provide changes to Assessment  uncertainty of this task.                          the quantified Fire PRA model. A conservative seismic hazard penalty is already applied to all RICT calculations to account for seismic risk impact.
Based on the discussion above, it is concluded that the methodology for the Seismic-Fire Interactions Assessment task does not introduce any epistemic uncertainties that affect the RICT calculation.
Therefore, RICT Program calculations are not impacted, and no RMAs are required to address this item.
E9-18
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty Table E9-3 ASSESSMENT OF FIRE PRA EPISTEMIC UNCERTAINTY Task
#  Description    Sources of Uncertainty                            Disposition for RICT Application 14  Fire Risk      As the culmination of other tasks, most of the    The selected truncation was confirmed to be consistent with the Quantification  uncertainty associated with quantification has    requirements of the PRA Standard (Reference 8).
already been addressed. The other source of uncertainty is the selection of the truncation    Based on a review of the assumptions and potential sources of limit. However, the selected truncation was      uncertainty related to this element and the discussion above, it is confirmed to be consistent with the              concluded that the methodology for the Fire Risk Quantification task requirements of the PRA Standard                  does not introduce any epistemic uncertainties that would affect the (Reference 8).                                    RICT calculation.
Therefore, RICT Program calculations are not impacted, and no RMAs are required to address this item.
15  Uncertainty    This task does not introduce any new              This task does not introduce any new uncertainties. This task is and Sensitivity uncertainties. This task is intended to          intended to address how the fire risk assessment could be impacted by Analyses        address how the fire risk assessment could        the various sources of uncertainty.
be impacted by the various sources of uncertainty.                                      Based on the discussion above, it is concluded that the methodology for the Uncertainty and Sensitivity Analyses task does not introduce any epistemic uncertainties that would affect the RICT calculation.
Therefore, RICT Program calculations are not impacted, and no RMAs are required to address this item.
16  FPRA            This task does not introduce any new              This task does not introduce any new uncertainties to the fire risk as it Documentation  uncertainties to the fire risk.                  outlines documentation requirements.
Based on the discussion above, it is concluded that the methodology for the FPRA documentation task does not introduce any epistemic uncertainties that would affect the RICT calculation.
Therefore, RICT Program calculations are not impacted, and no RMAs are required to address this item.
E9-19
 
ENCLOSURE 9 Key Assumptions and Sources of Uncertainty
: 5. 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, October, 2012 (ADAMS Accession No. ML12286A322).
: 2. NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, Revision 1, March 2017.
: 3. EPRI 1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, December 2008.
: 4. CL-PRA-013, Clinton Power Station Probabilistic Risk Assessment Summary Notebook, Rev. 6, February 2020.
: 5. EPRI 1026511, Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncertainty, December 2012.
: 6. CL-PRA-021.12, Clinton Power Station Fire PRA Uncertainty and Sensitivity Analysis Notebook, Rev. 2, February 2020.
: 7. CL-MISC-033, Assessment of Key Assumptions and Sources of Uncertainty for Risk-Informed Applications, Rev. 0, March 2020.
: 8. 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.
: 9. NUREG/CR-6595, An Approach for Estimating the Frequencies of Various Containment Failure Modes and Bypass Events, Rev. 1, October 2004.
: 10. NUREG/CR-6850 (also EPRI 1011989), "Fire PRA Methodology for Nuclear Power Facilities," September 2005, with Supplement 1 (EPRI 1019259), September 2010.
: 11. NUREG/CR-7150, Joint Asessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE), October 2012.
: 12. NUREG-2169 / EPRI 300200936, Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database, Rev. 0, January 2015.
E9-20
 
ENCLOSURE 10 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Program Implementation
 
ENCLOSURE 10 Program Implementation
: 1. Introduction Section 4.0, Item 11 of the NRC Final Safety Evaluation (Reference 1) for NEI 06-09 (Reference 2) requires that the license amendment request (LAR) provide a description of the implementing programs and procedures regarding the 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).
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).
: 2. RICT Program and Procedures Exelon will develop a program description and implementing procedures for the RICT Program.
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 included in NEI 06-09.
The program will be integrated with the online work control process. The work control process currently identifies the need to enter an LCO Action statement as part of the planning process and will additionally identify whether the provisions of the RICT program are required for the planned work. The risk thresholds associated with 10 CFR 50.65(a)(4) will be coordinated with the RICT limits. The Maintenance Rule program and Mitigating System Performance Index (MSPI) thresholds will assist in controlling the amount of risk expended in use of the RICT program.
The Operations Department (licensed operators) is responsible for compliance with the TS and will be responsible for implementation of RICTs and RMAs. Entry into the RICT program will require management approval prior to pre-planned activities and as soon as practicable following emergent conditions.
The procedures for the RICT program will address the following attributes consistent with NEI 06-09:
* Plant management positions with authority to approve entry into the RICT Program.
* Important definitions related to the RICT Program.
* Departmental and position responsibilities for activities in the RICT Program.
* Plant conditions for which the RICT Program is applicable.
* Limitations on implementing RICTs under voluntary and emergent conditions.
* Implementation of the RICT Program 30-day back stop limit.
* Use of the Real-Time Risk tool.
* Guidance on recalculating RICT and risk management action time (RMAT) within 12 hours or within the most limiting front-stop CT after a plant configuration change.
E10-1
 
ENCLOSURE 10 Program Implementation
* Requirements to identify and implement RMAs when the RMAT is exceeded or is anticipated to be exceeded, and to consider common cause failure potential in emergent RICTs.
* Guidance on the use of RMAs including the conditions under which they may be credited in RICT calculations.
* Conditions for exiting a RICT.
* Requirements for training on the RICT Program.
* Documentation requirements related to individual RICT evaluations, implementation of extended CTs, and accumulated annual risk.
: 3. RICT Program Training The scope of training for the RICT Program will include rules for the new TS program, Real-Time Risk tool software, TS Actions included in the program, and procedures. This training will be conducted for the following Exelon personnel:
Site Personnel
* Operations Director
* Operations Personnel (Licensed and Non-Licensed)
* Operations Training
* Outage Manager
* On-line Manager
* Planning and Scheduling Personnel
* Work Week Managers
* Regulatory Assurance Personnel
* Selected Maintenance Personnel
* Engineering
* Risk Management
* Other Selected Management Corporate Personnel
* Operations Corporate Functional Area Manager
* Fleet Outages Corporate Functional Area Manager
* Licensing Management and Personnel
* Risk Management Personnel and Managers
* Training Management and Personnel
* Other Selected Management Training will be carried out in accordance with Exelon training procedures and processes. These procedures were written based on the Institute of Nuclear Power Operations (INPO)
Accreditation (ACAD) requirements, as developed and maintained by the National Academy for Nuclear Training. Exelon has planned three levels of training for implementation of the RICT Program. They are described below:
E10-2
 
ENCLOSURE 10 Program Implementation Level 1 Training This is the most detailed training. It is intended for the individuals who will be directly involved in the implementation of the RICT Program. This level of training includes the following attributes:
* Specific training on the revised TS
* Record keeping requirements
* Case studies
* Hands-on experience with the Real-Time Risk tool for calculating RMAT and RICT
* Identifying appropriate RMAs
* Common cause failure RMA considerations in emergent RICTs
* Other detailed aspects of the RICT Program Level 2 Training This training is applicable to plant management positions with authority to approve entry into the RICT Program, as well as supervisors, managers, and other personnel who will closely support RICT implementation. These individuals need a broad understanding of the purpose, concepts, and limitations of the RICT Program. Level 2 training is significantly more detailed than Level 3 training (described below), but it is different from Level 1 training in that hands-on time with the Real-Time Risk tool, case studies, and other specifics are not required.
Level 3 Training This training is intended for the remaining personnel who require an awareness of the RICT Program. These employees need basic knowledge of the RICT Program requirements and procedures. This training will cover the RICT Program concepts that are important to disseminate throughout the organization.
: 4. References
: 1. Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,'"
dated May 17, 2007 (ADAMS Accession No. ML071200238).
: 2. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"
Revision 0-A, dated October 12, 2012 (ADAMS Accession No. ML12286A322).
E10-3
 
ENCLOSURE 11 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" Monitoring Program
 
ENCLOSURE 11 Monitoring Program
: 1. Introduction Section 4.0, Item 12 of the NRC Final Safety Evaluation (Reference 1) for NEI 06-09 (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 (Reference 2). (Note that RG 1.174, Revision 2 [Reference 4], issued by the NRC in May 2011, made editorial changes to the applicable section referenced in the NRC safety evaluation for Section 4.0, Item 12.)
This enclosure provides a description of the process applied to monitor the cumulative risk impact of implementation of the Risk-Informed Completion Time (RICT) Program, specifically the calculation of cumulative risk of extended Completion Times (CTs). 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, Risk Informed Technical Specifications Initiative 4b (Reference 2).
General requirements for a Performance Monitoring Program for risk-informed applications are discussed in Element 3 of Regulatory Guide 1.174 (Reference 3).
: 2. Description of Monitoring Program The RICT Program will require calculation of cumulative risk impact at least every refueling cycle, not to exceed 24 months, consistent with the guidance in NEI 06-09 (Reference 2). For the assessment period under evaluation, data will be collected for the risk increase associated with each application of an extended CT for both core damage frequency (CDF) and large early release frequency (LERF), and the total risk will be calculated by summing all risk associated with each RICT application. This summation is the change in CDF or LERF above the zero maintenance baseline levels during the period of operation in the extended CT (i.e., beyond the front-stop CT). The change in risk will be converted to average annual values.
The total average annual change in risk for extended CTs will be compared to the guidance of RG 1.174, Figures 4 and 5 (Reference 4), acceptance guidelines for CDF and LERF, respectively. If the actual annual risk increase is acceptable (i.e., not in Region I of Figures 4 and 5 of RG 1.174), then RICT program implementation is acceptable for the assessment period. Otherwise, further assessment of the cause of exceeding the acceptance guidelines of RG 1.174 and implementation of any necessary corrective actions to ensure future plant operation is within the guidelines will be conducted under the corrective action program.
The evaluation of cumulative risk will also identify areas for consideration, such as:
* RICT applications that dominated the risk increase
* Risk contributions from planned vs. emergent RICT applications
* Risk Management Actions (RMAs) implemented but not credited in the risk calculations
* Risk impact from applying RICT to avoid multiple shorter duration outages
* Any specific RICT application that incurred a large proportion of the risk E11-1
 
ENCLOSURE 11 Monitoring Program Based on a review of the considerations above, corrective actions will be developed and implemented as appropriate. These actions may include:
* Administrative restrictions on the use of RICTs for specific high-risk configurations
* Additional RMAs for specific configurations
* Rescheduling planned maintenance activities
* Deferring planned maintenance to shutdown conditions
* Use of temporary equipment to replace out-of-service systems, structures, or components (SSC)
* Plant modifications to reduce risk impact of future planned maintenance configurations In addition to impacting cumulative risk, implementation of the RICT Program may potentially impact the unavailability of SSCs. The existing Maintenance Rule (MR) monitoring programs under 10 CFR 50.65(a)(1) and (a)(2) provide for evaluation and disposition of unavailability impacts which may be incurred from implementation of the RICT Program. The SSCs in the scope of the RICT Program are also in the scope of the MR, which allows the use of the MR Program.
The monitoring program for the MR, along with the specific assessment of cumulative risk impact described above, serve as the Implementation and Monitoring Program for the RICT Program as described in Element 3 of RG 1.174 (Reference 3) and NEI 06-09 (Reference 2).
: 3. References
: 1. Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, 'Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,'"
dated May 17, 2007 (ADAMS Accession No. ML071200238).
: 2. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"
Revision 0-A, October 2012 (ADAMS Accession No. ML12286A322).
: 3. Regulatory Guide 1.174, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 1, November 2002.
: 4. Regulatory Guide 1.174, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, May 2011.
E11-2
 
ENCLOSURE 12 License Amendment Request Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 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
 
ENCLOSURE 12 Risk Management Action Examples
: 1. Introduction This enclosure describes the process for identification and implementation of Risk Management Actions (RMA) applicable during extended Completion Times (CT) and provides examples of RMAs. RMAs will be governed by plant procedures for planning and scheduling maintenance activities. The procedures will provide guidance for the determination and implementation of RMAs when entering the Risk-Informed Completion Time (RICT) Program consistent with the guidance provided in NEI 06-09-A, Revision 0 (Reference 1).
: 2. Responsibilities For planned entries into the RICT Program, Work Management is responsible for developing the RMAs with assistance from Operations and Risk Management. Operations is responsible for approval and implementation of RMAs. For emergent entry into extended CTs, Operations is also responsible for developing the RMAs.
: 3. Procedural Guidance For planned maintenance activities, implementation of RMAs will be required if it is anticipated that the risk management action time (RMAT) will be exceeded. For emergent activities, RMAs must be implemented if the RMAT is reached. Also, if an emergent event occurs requiring recalculation of a RMAT already in place, the procedure will require a reevaluation of the existing RMAs for the new plant configuration to determine if new RMAs are appropriate. These requirements of the RICT Program are consistent with the guidance of NEI 06-09-A.
For emergent entry into a RICT, if the extent of condition is not known, RMAs related to the success of redundant and diverse SSCs and reducing the likelihood of initiating events relying on the affected function will be developed to address the increased likelihood of a common cause event.
RMAs will be implemented in accordance with current procedures (e.g., References 2, 3, 4,5) no later than the time at which an incremental core damage probability (ICDP) of 1E-6 is reached, or no later than the time when an incremental large early release probability (ILERP) of 1E-7 is reached. If, as the result of an emergent condition, the instantaneous core damage frequency (ICDF) or the instantaneous large early release frequency (ILERF) exceeds 1E-3 per year or 1E-4 per year, respectively, RMAs are also required to be implemented. These requirements are consistent with the guidelines of NEI 06-09-A.
By determining which structures, systems, or components (SSCs) are most important from a CDF or LERF perspective for a specific plant configuration, RMAs may be created to protect these SSCs. Similarly, knowledge of the initiating event or sequence contribution to the configuration-specific CDF or LERF allows development of RMAs that enhance the capability to mitigate such events. The guidance in NUREG-1855 (Reference 6) and EPRI TR-1026511 (Reference 7) 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, or important support systems or human E12-1
 
ENCLOSURE 12 Risk Management Action Examples actions). Enclosure 9 identifies several areas of uncertainty in the internal events and fire PRAs that will be considered in defining configuration-specific RMAs when entering a RICT.
If the planned activity or emergent condition includes an SSC that is identified to impact Fire PRA, as identified in the current Real Time Risk Program, Fire PRA specific RMAs associated with that SSC will be implemented per the current plant procedure.
It is possible to credit RMAs in RICT calculations, to the extent the associated plant equipment and operator actions are modeled in the PRA; however, such quantification of RMAs is neither required nor expected by NEI 06-09-A. Nonetheless, if RMAs will be credited to determine RICTs, the procedure instructions will be consistent with the guidance in NEI 06-09-A.
NEI 06-09-A classifies RMAs into the three categories described below:
: 1) Actions to increase risk awareness and control.
* Shift brief
* Pre-job brief
* Training
* Presence of system engineer or other expertise related to the activity
* Special purpose procedure to identify risk sources and contingency plans
: 2) Actions to reduce the duration of maintenance activities.
* Pre-staging materials
* Conducting training on mock-ups
* Performing the activity around the clock
* Performing walk-downs on the actual system(s) to be worked on prior to beginning work
: 3) Actions to minimize the magnitude of the risk increase.
* Suspend or minimize activities on redundant systems
* Suspend or minimize activities on other systems that adversely affect the CDF or LERF
* Suspend or minimize activities on systems that may cause a trip or transient to minimize the likelihood of an initiating event that the out-of-service component is meant to mitigate
* Use temporary equipment to provide backup power, ventilation, etc.
* Reschedule other risk-significant activities Determination of RMAs involves the use of both qualitative and quantitative considerations for the specific plant configuration and the practical means available to manage risk. The scope and number of RMAs developed and implemented are reached in a graded manner.
Procedural guidance for development of RMAs in support of the RICT program builds off the RMAs developed for other processes, such as the RMAs developed under the 10 CFR 50.65(a)(4) program and the protected equipment program. Additionally, Common Cause RMAs are developed to address the potential impact of common cause failures.
E12-2
 
ENCLOSURE 12 Risk Management Action Examples General RMAs are developed for input into the RICT system guidelines. These guidelines are listed in site-specific T&RMs and are developed using a graded approach. Consideration is given for system functionality and includes consideration for common cause impacts within the system. These RMAs include:
* Consideration of rescheduling maintenance to reduce risk
* Discussion of RICT in pre-job briefs
* Consideration of proactive return-to-service of other equipment
* Efficient execution of maintenance In addition to the RMAs developed qualitatively for the system guidelines, RMAs are developed based on the Real-Time Risk tool to identify configuration-specific RMA candidates to manage the risk associated with internal events, internal flooding, and fire events. These actions include:
* Identification of important equipment or trains for protection
* Identification of important Operator Actions for briefings
* Identification of key fire initiators and fire zones for RMAs in accordance with the site Fire RMA process
* Identification of dominant initiating events and actions to minimize potential for initiators
* Consideration of insights from PRA model cutsets, through comparison of importances Common cause RMAs are also developed to ensure availability of redundant SSCs, to ensure availability of diverse or alternate systems, to reduce the likelihood of initiating events that require operation of the out-of-service components, and to prepare plant personnel to respond to additional failures. Common cause RMAs are developed by considering the impact of loss of function for the affected SSCs.
Examples of common cause RMAs include:
* Performance of non-intrusive inspections on alternate trains
* Confidence runs performed for standby SSCs
* Increased monitoring for running components
* Expansion of monitoring for running components
* Deferring maintenance and testing activities that could generate an initiating event which would require operation of potentially affected SSCs
* Readiness of operators and maintenance to respond to additional failures
* Shift briefs or standing orders which focus on initiating event response or loss of potentially affected SSCs Per Exelon procedure, for emergent conditions where the extent of condition is not performed prior to entering into the Risk Management Action Times or the extent of condition cannot rule out the potential for common cause failure, common cause RMAs are expected to be implemented to mitigate common cause failure potential and impact. These can include the pre-identified RMAs included in the system guidelines as discussed above, as well as alternative common cause RMAs for the specific configuration. Alternate RMAs, including both E12-3
 
ENCLOSURE 12 Risk Management Action Examples regular and common cause considerations, are developed for the specific configuration following the steps outlined above.
: 4. 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 unavailability of one Diesel Generator (DG), one Offsite Circuit, one Diesel Generator (DG) and one Offsite Circuit, one Battery Charger, or one Residual Heat Removal (RHR) pump.
4.1. Electrical Action Statements 4.1.1  For TS action 3.8.1.B, One required Division DG inoperable. Additional RMAs would include:
: 1. Actions to increase risk awareness and control.
* Briefing of the on-shift Operations crew concerning the unit activities, including any compensatory measures established, and review of the appropriate emergency operating procedures for a Loss of Offsite Power and station blackout including bus crossties.
* Notification of the TSO of the configuration so that any planned activities with the potential to cause a grid disturbance are deferred.
* Proactive implementation of RMAs during times of high grid stress conditions, such as during high demand conditions.
: 2. Actions to reduce the duration of maintenance activities.
* For preplanned RICT entry, creation of a sub schedule related to the specific evolution which is reviewed for personnel resource availability.
* Confirmation of parts availability prior to entry into a preplanned RICT.
* Walkdown of work prior to execution.
: 3. Actions to minimize the magnitude of the risk increase.
* Evaluation of weather conditions for threats to the reliability of offsite power supplies.
* Deferral of elective maintenance in the switchyard, on the station electrical distribution systems, and on the main and auxiliary transformers associated with the unit.
* Deferral of planned maintenance or testing that affects the reliability of operable DGs and their associated support equipment. Treat the remaining operable DGs as protected equipment.
E12-4
 
ENCLOSURE 12 Risk Management Action Examples
* Deferral of planned maintenance or testing on redundant train safety systems. If testing or maintenance activities must be performed, a review of the potential risk impact will be performed.
* Implementation of 10 CFR 50.65(a)(4) fire-specific RMAs associated with the affected DG.
4.1.2  TS action 3.8.1.A, one offsite circuit inoperable, additional RMAs would include:
: 1. Actions to increase risk awareness and control.
* Briefing of the on-shift operations crew concerning the unit activities, including any compensatory measures established, and review of the appropriate emergency operating procedures for a Loss of Offsite Power and station blackout including bus crossties.
* Notification of the TSO of the configuration so that any planned activities with the potential to cause a grid disturbance are deferred.
* Proactive implementation of RMAs during times of high grid stress conditions prior to reaching the RMAT, such as during high demand conditions.
: 2. Actions to reduce the duration of maintenance activities.
* For preplanned RICT entry, creation of a sub schedule related to the specific evolution which is reviewed for personnel resource availability
* Confirmation of parts availability prior to entry into a preplanned RICT.
* Walkdown of work prior to execution.
: 3. Actions to minimize the magnitude of the risk increase.
* Evaluation of weather conditions for threats to the reliability of remaining offsite power supplies.
* Deferral of elective maintenance in the switchyard, on the station electrical distribution systems, and on the main and auxiliary transformers associated with the unit.
* Protection of the remaining offsite sources, including switchyard and transformers.
* Deferral of planned maintenance or testing that affects the reliability of DGs and their associated support equipment. Treat these as protected equipment.
* Implementation of 10 CFR 50.65(a)(4) fire-specific RMAs associated with the affected offsite source.
E12-5
 
ENCLOSURE 12 Risk Management Action Examples 4.1.3  TS action 3.8.1.D, one required offsite circuit inoperable AND One required DG inoperable, additional RMAs would include:
: 1. Actions to increase risk awareness and control.
* Briefing of the on-shift operations crew concerning the unit activities, including any compensatory measures established, and review of the appropriate emergency operating procedures for a Loss of Offsite Power and station blackout.
* Notification of the TSO of the configuration so that any planned activities with the potential to cause a grid disturbance are deferred.
* Proactive implementation of RMAs during times of high grid stress conditions prior to reaching the RMAT, such as during high demand conditions.
* For a planned RICT, prior to removal from service the actions in the associated loss of bus procedure would be reviewed and implemented.
: 2. Actions to reduce the duration of maintenance activities.
* For preplanned RICT entry, creation of a sub schedule related to the specific evolution which is reviewed for personnel resource availability.
* Confirmation of parts availability prior to entry into a preplanned RICT.
* Walkdown of work prior to execution.
: 3. Actions to minimize the magnitude of the risk increase.
* Deferral of elective maintenance in the switchyard, on the station electrical distribution systems, and on the main and auxiliary transformers associated with the unit.
* Deferral of planned maintenance or testing that affects the reliability of DGs and their associated support equipment for the remaining buses.
* Implementation of 10 CFR 50.65(a)(4) fire-specific RMAs associated with the affected bus.
4.1.4  TS action 3.8.4.A, One required DC battery charger on Division 1 or 2 inoperable.
Additional RMAs would include:
: 1. Actions to increase risk awareness and control.
* Briefing of the on-shift operations crew concerning the unit activities, including any compensatory measures established, and review of the appropriate emergency operating procedures for a Loss of DC division and station blackout.
* Briefing of the on-shift operations crew concerning the impact the DC division has on the potential response to plant events such as reduced control systems.
* Prior to removal from service. If a Planned RICT, the actions in the associated loss of bus procedure would be reviewed and implemented.
E12-6
 
ENCLOSURE 12 Risk Management Action Examples
* Minimize activities that could trip the unit.
: 2. Actions to reduce the duration of maintenance activities.
* For preplanned RICT entry, creation of a sub schedule related to the specific evolution which is reviewed for personnel resource availability.
* Confirmation of parts availability prior to entry into a preplanned RICT.
* Walkdown of work prior to execution.
: 3. Actions to minimize the magnitude of the risk increase.
* Deferral of elective maintenance in the switchyard, on the station electrical distribution systems, and on the main and auxiliary transformers associated with the unit.
* Protection of the remaining DC electrical buses in that unit.
* Remove nonessential loads from battery to extend time voltage will remain above minimum required level.
* Implementation of 10 CFR 50.65(a)(4) fire-specific RMAs associated with the affected bus.
4.2. ECCS Action Statements 4.2.1  TS Action 3.5.1.A, one low pressure ECCS injection/spray subsystem inoperable, the RMAs would include the following.
: 1. Actions to reduce the duration of maintenance activities.
* For preplanned RICT entry, creation of a sub schedule related to the specific evolution which is reviewed for personnel resource availability
* Confirmation of parts availability prior to entry into a preplanned RICT.
* Walkdown of work prior to execution.
: 2. Actions to minimize the magnitude of the risk increase.
* Verify system alignment of low pressure ECCS.
* Defer planned maintenance or testing that affects the reliability of those safety systems that provide a defense-in-depth. If testing or maintenance activities must be performed, a review of the potential risk impact will be performed.
* Minimize activities that could trip the unit.
* Defer planned maintenance or testing activities on the redundant ECCS loops and its associated support equipment. Treat those systems as protected equipment.
E12-7
 
ENCLOSURE 12 Risk Management Action Examples
* Implement 10 CFR 50.65(a)(4) fire-specific RMAs associated with the affected ECCS loop.
: 5. 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)
: 2. Exelon Procedure OP-AA-201-012-1001, "Operations On-Line Fire Risk Management"
: 3. CL-CRM-014, "Development of Risk Management Actions for the Inclusion of Fire Insights into Clinton Power Station Configuration Risk Management Program"
: 4. Exelon Procedure WC-AA-101-1006, "On-Line Risk Management and Assessment"
: 5. Exelon Procedure OP-AA-108-117, "Protected Equipment Program"
: 6. NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," U.S. Nuclear Regulatory Commission, March 2009.
: 7. EPRI TR-1026511, "Practical Guidance on the Use of Probabilistic Risk Assessment in Risk- Informed Applications with a Focus on the Treatment of Uncertainty," Technical Update, Electric Power Research Institute, December 2012 E12-8}}

Latest revision as of 21:10, 12 December 2024

Application to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Extended Completed Times-RITSTF Initiative 4b
ML20121A178
Person / Time
Site: Clinton Constellation icon.png
Issue date: 04/30/2020
From: Simpson P
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-20-052
Download: ML20121A178 (373)
v • d • e


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