ML15166A075
| ML15166A075 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/04/2015 |
| From: | Jim Barstow Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TSTF-425, Rev. 3 | |
| Download: ML15166A075 (253) | |
Text
l&Exelon Generation 200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90 June 4, 2015 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 R.E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244
SUBJECT:
Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR 50.90), "Application for amendment of license, construction permit, or early site permit,"
Exelon Generation Company, LLC (Exelon) is submitting a request for an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No.
DPR-18 for R. E. Ginna Nuclear Power Plant (Ginna).
The proposed amendment would modify Ginna's TS by relocating specific surveillance frequencies to a licensee-controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies."
The changes are consistent with NRC-approved Industry Technical Specifications Task Force (TSTF) Standard Technical Specifications (STS) change TSTF-425, "Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force (RITSTF)
Initiative 5b, Revision 3," (ADAMS Accession No. ML090850642). The Federal Register Notice published on July 6, 2009 (74 FR 31996), announced the availability of this TS improvement.
Attachment 1 provides a description of the proposed change, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides documentation of Probabilistic Risk Assessment (PRA) technical adequacy. Attachment 3 provides the existing Ginna TS pages marked up to show the proposed changes. Attachment 4 provides the proposed Ginna TS Bases changes. Attachment 5 provides a TSTF-425 versus Ginna TS Cross-Reference. Attachment 6 provides the proposed No Significant Hazards Consideration. Attachment 7 provides the proposed inserts. A-U cii
License Amendment Request Adoption of TSTF-425, Rev. 3 Docket No. 50-244 June 4, 2015 Page 2 There are no regulatory commitments contained in this letter.
Exelon requests approval of the proposed license amendment by June 4, 2016, with the amendment being implemented within 120 days.
These proposed changes have been reviewed by the Plant Operations Review Committee and approved in accordance with Nuclear Safety Review Board procedures.
In accordance with 10 CFR 50.91, "Notice for Public Comment; State Consultation," a copy of this application, with attachments, is being provided to the designated State Official.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 4 th day of June 2015.
If you should have any questions regarding this submittal, please contact Enrique Villar at 610-765-5736.
Respectfully, James Barstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Description and Assessment
- 2. Documentation of PRA Technical Adequacy
- 3. Proposed Technical Specification Page Changes
- 4. Proposed Technical Specification Bases Page Changes
- 5. TSTF-425 (NUREG-1431) vs. Ginna Cross-Reference
- 6. Proposed No Significant Hazards Consideration
- 7. Proposed Inserts cc: USNRC Region I Regional Administrator w/attachments I1 USNRC Senior Resident Inspector - Ginna USNRC Project Manager, NRR - Ginna A. L. Peterson, NYSERDA
ATTACHMENT 1 License Amendment Request R. E. Ginna Nuclear Power Plant Docket No. 50-244 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
Description and Assessment
LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-244 Page 1 of 5 DESCRIPTION AND ASSESSMENT
1.0 DESCRIPTION
The proposed amendment would modify the R. E. Ginna Nuclear Power plant (Ginna) Technical Specifications (TS) by relocating specific TS surveillance frequencies to a licensee-controlled program with the adoption of Technical Specification Task Force (TSTF) - 425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b" (Ref. 1). Additionally, the change would add a new program, the Surveillance Frequency Control Program, to TS Section 5, Administrative Controls.
The changes are consistent with NRC-approved Industry/lTSTF Standard Technical Specifications (STS) change TSTF-425, Revision 3, (ADAMS Accession No. ML090850642). The Federal Register notice published on July 6, 2009 (74 FR 31996) (Ref. 2), announced the availability of this TS improvement.
2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation Exelon Generation Company, LLC (Exelon) has reviewed the NRC staff's Model Safety Evaluation for TSTF-425, Revision 3, dated July 6, 2009. This review included a review of the NRC staff's Model Safety Evaluation, TSTF-425, Revision 3, and the requirements specified in NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. ML071360456) (Ref.
3).
The traveler and Model Safety Evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). Ginna was not licensed to the 10 CFR 50, Appendix A GDC. However, the Ginna's Updated Final Safety Analysis Report (UFSAR), in Section 3.1 "Conformance with NRC General Design Criteria,"
provides an assessment against the GDC. Based on the assessment performed and described in the in the Ginna UFSAR, Exelon believes that the plant-specific requirements for Ginna are sufficiently similar to the Appendix A GDC and represent an adequate technical basis for adopting the proposed change.
Attachment 2 includes Exelon's documentation with regard to Probabilistic Risk Assessment (PRA) technical adequacy consistent with the requirements of Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," (ADAMS Accession No. ML070240001) (Ref. 4), Section 4.2, and describes any PRA models without NRC-endorsed standards, including documentation of the quality characteristics of those models in accordance with Regulatory Guide 1.200.
Exelon has concluded that the justifications presented in the TSTF proposal and the NRC staff's Model Safety Evaluation prepared by the NRC staff are applicable to Ginna and justify this amendment to incorporate the changes to the Ginna TS.
LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-244 Page 2 of 5 2.2 Optional Changes and Variations The proposed amendment is consistent with the STS changes described in TSTF-425, Revision 3; however, Exelon proposes variations or deviations from TSTF-425, as identified below, which includes differing Surveillance numbers.
Revised (clean) TS pages are not included in this amendment request given the number of TS pages affected, the straightforward nature of the proposed changes, and outstanding Ginna amendment requests that will impact some of the same TS pages. 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," (Ref. 5) in that the mark-ups fully describe the changes desired. This is an administrative deviation from the NRC staff's model application dated July 6, 2009 (74 FR 31996) with no impact on the NRC staff's Model Safety Evaluation published in the same Federal Register Notice. As a result of this deviation, the contents and numbering of the attachments for this amendment request differ from the attachments specified in the NRC staff's model application After NRC approval of TSTF-425, it was recognized that surveillance frequencies that have not been changed under the Surveillance Frequency Control Program (SFCP) may not be based on operating experience, equipment reliability or plant risk. Therefore, the TSTF and the NRC agreed that the TSTF-425 TS Bases insert, "The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program," should be revised to state, "The Surveillance Frequency is controlled under the Surveillance Frequency Control Program." The existing TS Bases information will be relocated to the licensee-controlled SFCP.
Attachment 5 provides a cross-reference between TSTF-425 versus the Ginna Surveillances included in this amendment request. Attachment 5 includes a summary description of the referenced TSTF-425 TS Surveillances, which is provided for information purposes only and is not intended to be a verbatim description of the TS Surveillances. This cross-reference highlights the following:
- a. Surveillances included in TSTF-425 and corresponding Ginna Surveillances have differing Surveillances numbers,
- b. Surveillances included in TSTF-425 that are not contained in the Ginna TS, and
- c. Ginna plant-specific Surveillances that are not contained in TSTF-425 Surveillances and, therefore, are not included in the TSTF-425 mark-ups.
In addition, there are Surveillances contained in TSTF-425 that are not contained in the Ginna TS. Therefore, the NUREG-1431 mark-ups included in TSTF-425 for these Surveillances are not applicable to Ginna. This is an administrative deviation from TSTF-425 with no impact on the NRC staff's Model Safety Evaluation dated July 6, 2009 (74 FR 31996).
Ginna TS include plant-specific Surveillances that are not contained in NUREG-1431 and, therefore, are not included in the NUREG-1431 mark-ups provided in TSTF-425.
Exelon has determined that the relocation of the Frequencies for these Ginna plant-specific Surveillances is consistent with TSTF-425, Revision 3, and with the NRC staff's
LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-244 Page 3 of 5 Model Safety Evaluation dated July 6, 2009 (74 FR 31996), including the scope exclusions identified in Section 1.0, "Introduction," of the Model Safety Evaluation.
Changes to the Frequencies for these plant-specific Surveillances would be controlled under the SFCP. The SFCP provides the necessary administrative controls to require that Surveillances related to testing, calibration and inspection are conducted at a frequency to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the Limiting Conditions for Operation will be met. Changes to Frequencies in the SFCP would be evaluated using the methodology and probabilistic risk guidelines contained in NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. ML071360456), as approved by NRC letter dated September 19, 2007 (ADAMS Accession No. ML072570267). The NEI 04-10, Revision 1 methodology includes qualitative considerations, risk analyses, sensitivity studies and bounding analyses, as necessary, and recommended monitoring of the performance of systems, components, and structures (SSCs) for which Frequencies are changed to assure that reduced testing does not adversely impact the SSCs. In addition, the NEI 04-10, Revision 1 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) (Ref. 6),
relative to changes in Surveillance Frequencies. Therefore, the proposed relocation of the Ginna plant-specific Surveillance Frequencies is consistent with TSTF-425 and with the NRC staff's Model Safety Evaluation dated July 6, 2009 (74 FR 31996).
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Exelon has reviewed the proposed no significant hazards consideration (NSHC) determination published in the Federal Register dated July 6, 2009 (74 FR 31996). Exelon has concluded that the proposed NSHC presented in the Federal Register notice is applicable to Ginna, and is provided as Attachment 6 to this amendment request, which satisfies the requirements of 10 CFR 50.91 (a), "Notice for public comment; State consultation" (Ref. 7).
3.2 Applicable Regulatory Requirements A description of the proposed changes and their relationship to applicable regulatory requirements is provided in TSTF-425, Revision 3 (ADAMS Accession No. ML090850642) and the NRC staff's Model Safety Evaluation published in the Notice of Availability dated July 6, 2009 (74 FR 31996). Exelon has concluded that the relationship of the proposed changes to the applicable regulatory requirements presented in the Federal Register notice is applicable to Ginna.
3.3 Precedence This application is being made in accordance with the TSTF-425, Revision 3 (ADAMS Accession No. ML090850642). Exelon is not proposing significant variations or deviations from the TS changes described in TSTF 425 or in the content of the NRC staff's Model Safety Evaluation published on July 6, 2009 (74 FR 31996). The NRC has previously approved
LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-244 Page 4 of 5 amendments to the TS as part of the pilot process for TSTF-425, including but not limited to Amendment Nos. 186 and 147 for Limerick Generating Station, Amendment No.276 for Oyster Creek Nuclear Power Station dated September 27, 2010; Amendment Nos. 200 and 201 for Diablo Canyon Power Plant, Units 1 and 2, respectively, dated October 30, 2008; and Amendment Nos. 188 and 175 for South Texas Project, Units 1 and 2, respectively, dated October 31, 2008. The subject License Amendment Request proposes to relocate periodic surveillance frequencies to a licensee-controlled program and add a new program (the Surveillance Frequency Control Program) to the Administrative Controls section of TS in accordance with TSTF-425 and as discussed in the previously approved amendments.
3.4 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
Exelon has reviewed the environmental consideration included in the NRC staff's Model Safety Evaluation published in the Federal Register on July 6, 2009 (74 FR 31996). Exelon has concluded that the staff's findings presented therein are applicable to Ginna, and the determination is hereby incorporated by reference for this application.
LAR - Adoption of TSTF-425, Revision 3 Attachment 1 Docket No. 50-244 Page 5 of 5
5.0 REFERENCES
- 1. TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b," March 18, 2009 (ADAMS Accession Number: ML090850642).
- 2. NRC Notice of Availability of Technical Specification Improvement to Relocate Surveillance Frequencies to Licensee Control - Risk-Informed Technical Specification Task Force (RITSTF) Initiative 5b, Technical Specification Task Force - 425, Revision 3, published on July 6, 2009 (74 FR 31996).
- 3. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," April 2007 (ADAMS Accession Number:
- 4. Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," January 2007 (ADAMS Accession Number: ML070240001).
- 5. 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit."
- 6. Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:
Technical Specifications," dated August 1998 (ADAMS Accession No. ML003740176).
- 7. 10 CFR 50.91(a), "Notice for public comment; State consultation."
ATTACHMENT 2 License Amendment Request R. E. Ginna Nuclear Power Plant Docket No. 50-244 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
Documentation of PRA Technical Adequacy
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page i of i Documentation of PRA Technical Adequacy TABLE OF CONTENTS Section Paqe 1 .0 O ve rv ie w ......................................................................................................................... 1 2.0 Technical Adequacy of the PRA Model ....................................................................... 2 2.0.1 PRA Maintenance and Update ....................................................................... 3 2.0.2 Plant Changes not yet Incorporated into the PRA Model ................................ 4 2.0.3 Applicability of Peer Review Findings and Observations .................................. 4 2.0.4 Consistency with Applicable PRA Standards .................................................. 5 2.0.5 Identification of Key Assumptions ................................................................... 5 2.1 External Events Considerations .................................................................................. 5 2 .2 S u m m a ry ......................................................................................................................... 7 2 .3 R efe re nce s ...................................................................................................................... 7
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 1 of 29 Documentation of PRA Technical Adequacy 1.0 Overview The implementation of the Surveillance Frequency Control Program (also referred to as Technical Specifications Initiative 5b) at the Ginna Nuclear Power Plant will follow the guidance provided in NEI 04-10, Revision 1 [Ref. 11 in evaluating proposed surveillance test interval (STI; also referred to as "surveillance frequency") changes.
The following steps of the risk-informed STI revision process are common to proposed changes to all STIs within the proposed licensee-controlled program.
- Each STI revision is reviewed to determine whether there are any commitments made to the NRC that may prohibit changing the interval. If there are no related commitments, or the commitments may be changed using a commitment change process based on NRC endorsed guidance, then evaluation of the STI revision would proceed. If a commitment exists and the commitment change process does not permit the change, then the STI revision would not be implemented.
" A qualitative analysis is performed for each STI revision that involves several considerations as explained in NEI 04-10, Revision 1.
- Each STI revision is reviewed by an Expert Panel, referred to as the Integrated Decision making Panel (IDP), which is normally the same panel as is used for Maintenance Rule implementation, but with the addition of specialists with experience in surveillance tests and system or component reliability. If the IDP approves the STI revision, the change is documented and implemented, and available for audit by the Nuclear Regulatory Commission (NRC). If the IDP does not approve the STI revision, the STI value is left unchanged.
" Performance monitoring is conducted as recommended by the IDP. In some cases, no additional monitoring may be necessary beyond that already conducted under the Maintenance Rule. The performance monitoring helps to confirm that no failure mechanisms related to the revised test interval become important enough to alter the information provided for the justification of the interval changes.
- The IDP is responsible for periodic review of performance monitoring results. If it is determined that the time interval between successive performances of a surveillance test is a factor in the unsatisfactory performances of the surveillance, the IDP returns the STI back to the previously acceptable STI.
- In addition to the above steps, the Probabilistic Risk Assessment (PRA) is used when possible to quantify the effect of a proposed individual STI revision compared to acceptance criteria in NEI 04-10. Also, the cumulative impact of all risk-informed STI revisions on all PRAs (i.e., internal events, external events and shutdown) is also compared to the risk acceptance criteria as delineated in NEI 04-10.
For those cases where the STI cannot be modeled in the plant PRA (or where a particular PRA model does not exist for a given hazard group), a qualitative or bounding analysis is performed to provide justification for the acceptability of the proposed test interval change.
The NEI 04-10 [Ref. 1 methodology endorses the guidance provided in Regulatory Guide (RG) 1.200, Revision 1 [Ref. 21, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities." The guidance in RG-1.200
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 2 of 29 indicates that the following steps should be followed when performing PRA assessments (NOTE: Because of the broad scope of potential Initiative 5b applications and the fact that the risk assessment details will differ from application to application, each of the issues encompassed in Items 1 through 3 below will be covered with the preparation of each individual PRA assessment made in support of the individual STI interval requests. Item 3 satisfies one of the requirements of Section 4.2 of RG 1.200. The remaining requirements of Section 4.2 are addressed by Item 4 below.):
- 1. Identify the parts of the PRA used to support the application
- SSCs, operational characteristics affected by the application and how these are implemented in the PRA model
- A definition of the acceptance criteria used for the application
- 2. Identify the scope of risk contributors addressed by the PRA model
- If not full scope (i.e., internal and external), identify appropriate compensatory measures or provide bounding arguments to address the risk contributors not addressed by the model.
- 3. Summarize the risk assessment methodology used to assess the risk of the application
- Include how the PRA model was modified to appropriately model the risk impact of the change request.
- 4. Demonstrate the Technical Adequacy of the PRA
- Identify plant changes (design or operational practices) that have been incorporated at the site, but are not yet in the PRA model and justify why the change does not impact the PRA results used to support the application.
- Document peer review findings and observations that are applicable to the parts of the PRA required for the application, and for those that have not yet been addressed justify why the significant contributors would not be impacted.
- Document that the parts of the PRA used in the decision are consistent with applicable standards endorsed by the Regulatory Guide ( RG-1.200 Revision 1 was used for the Ginna Internal Events PRA Peer review). Provide justification to show that where specific requirements in the standard are not adequately met, it will not unduly impact the results.
- Identify key assumptions and approximations relevant to the results used in the decision-making process.
The purpose of the remaining portion of this attachment is to address the requirements identified in Item 4 above.
2.0 Technical Adequacy of the PRA Model The GN1 14A-W version of the Ginna PRA model is the most recent evaluation of internal event risks. The Ginna PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA model quantification process used for the Ginna PRA is based on the event tree / fault tree methodology, which is a well-known methodology in the industry.
Exelon Generation Company, LLC (Exelon) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 3 of 29 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. Prior to joining the Exelon nuclear fleet in 2014, comparable practices were in place when Ginna was owned and operated by Constellation Energy Nuclear Group (CENG).
Because of the similarities between the CENG and Exelon practices, no additional discussion specifically regarding the legacy CENG approach will be provided. The following information describes the Exelon approach (and by extension the CENG approach) to PRA model maintenance, as it applies to the Ginna PRA.
2.0.1 PRA Maintenance and Update The Exelon risk management process ensures that the applicable PRA model is an accurate reflection of the as-built and as-operated plants. This process is defined in the Exelon Risk Management program, which consists of a governing procedure (ER-AA-600, "Risk Management") and subordinate implementation training and reference materials (T&RM's).
- Exelon T&RM ER-AA-600-1015, "Full Power Internal Event (FPIE) PRA Model Update,"
delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating Exelon nuclear generation sites.
- ER-AA-600-1061 "Fire PRA Model Update and Control" delineates the responsibilities and guidelines for updating the station fire PRA.
The overall Exelon Risk Management program, including ER-AA-600-1015 and ER-AA-600-1061, define the process for: implementing regularly scheduled and interim PRA model updates; for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, industry operating experience, etc.); and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plants, the following activities are routinely performed:
" Design changes and procedure changes are reviewed for their impact on the PRA model.
- New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.
- Maintenance unavailabilities are captured, and their impact on CDF is trended.
- Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated during each model update.
In addition to these activities, Exelon risk management procedures provide the guidance for particular risk management maintenance activities. This guidance includes:
" The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications.
- Guidelines for updating the full power, internal events PRA models for Exelon nuclear generation sites.
" Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 4 of 29 modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (10 CFR 50.65(a)(4)).
An application specific update of the PRA model was completed in the 4th quarter of 2014 to support an update of the Mitigating System Performance Indicator (MSPI) application. Exelon will be performing a Full Power Internal Events (FPIE) model update to the Ginna PRA in 2015.
As indicated previously, RG 1.200 also requires that additional information be provided as part of the LAR submittal to demonstrate the technical adequacy of the PRA model used for the risk assessment. Each of these items (plant changes not yet incorporated in to the PRA model, relevant peer review findings, consistency with applicable PRA Standards, and the identification of key assumptions) will be discussed in turn.
2.0.2 Plant Changes Not Yet Incorporated into the PRA Model Each Exelon station maintains an updating requirements evaluation (URE) database to track all enhancements, corrections, and unincorporated plant changes. During the normal screening conducted as part of the plant change process, if a potential model update is identified a new URE database item is created. Depending on the potential impact of the identified change, the requirements for incorporation will vary.
As part of the PRA evaluation for each STI change request, a review of open items in the URE database for GINNA will be performed and an assessment of the impact on the results of the application will be made prior to presenting the results of the risk analysis to the IDP. If a non-trivial impact is expected, then this may include the performance of additional sensitivity studies or PRA model changes to confirm the impact on the risk analysis.
2.0.3 Applicability of Peer Review Findings and Observations A PRA model update was completed in 2009, resulting in the Ginna PRA Model 6.5. The Ginna PRA model was revised to meet RG 1.200, revision 1, guidance and comply with the ASME/ANS PRA Standard RA-Sc-2007[Ref. 3.
This model was peer reviewed under the auspices of the PWR Owners Group (PWROG) in the 2nd quarter of 2009 [Ref. 71. This peer review was performed following NEI 05-04 [Ref. 51, and NEI 00-002 [Ref. 61. This peer review included an assessment of the PRA model maintenance and update process.
Since the 2008 peer review, an application specific PRA model update was completed in 2012 to support implementation of NFPA-805. As part of the development of this model a peer review of the fire PRA was conducted in June of 2012 [Ref. 8]. This peer review used NEI 12 [Ref. 9] to evaluate the model against the ASME PRA Standard (ASME/ANS RA-Sa-2009)
[Ref. 10] along with the NRC clarifications provided in Regulatory Guide 1.200, Rev. 2 [Ref. 22].
Since the 2012 peer review, several updates to the Ginna PRA have taken place. An application specific model update was completed in December 2014 to support the Mitigating System Performance Indicator (MSPI) process. The latest model is the GN1 14A-W. This model includes the addition of two diesel generators for providing an alternate source of power to the Standby AFW (SAFW) Pumps and a condensate storage tank to provide a dedicated source of water to the SAFW Pumps.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 5 of 29 2.0.4 Consistency with Applicable PRA Standards As indicated above there have been two relevant peer reviews conducted on the current PRA model.
- The 2009 peer review for the PRA ASME model update identified 309 Supporting Requirements (SR) applicable to the Ginna PRA. Of these 29 were not met, 2 met capability category (cc) 1, 13 partially met cc 2, 31 met cc 2, 22 partially met cc 3, 14 met cc 3, and 198 fully met all capability requirements. There were 24 findings and observations (F&O's) issued to address the identified gaps to compliance with the PRA standard. Subsequent to the peer review, 13 of the findings have been addressed and 11 are still open pending the next model update. The F&O's are listed in Table 2-1 which includes what, if any impact, there may be to the assessment of STIs for the 5b initiative.
- The 2012 fire PRA peer review for the PRA ASME model update identified 183 Supporting Requirements (SR) to be reviewed for the Ginna PRA. Of these 2 were not met, 2 met capability category (cc) 1, 8 partially met cc 2, 17 met cc 2, 13 partially met cc 3, 7 met cc 3, and 118 fully met all capability requirements and 16 were not applicable. There were 19 findings and 22 suggestions issued to address potential gaps to compliance with the PRA standard. There were 3 Best Practices. All of the findings from the fire PRA peer review have since been closed. As the results of this peer review have already been communicated to the NRC as part of the NFPA-805 submittal [Ref.
12] and subsequent requests for additional information (RAI), these will not be catalogued in this document.
All remaining gaps will be reviewed for consideration during the 2015 model update but are judged to have low impact on the PRA model or its ability to support a full range of PRA applications. The remaining gaps are documented in the URE database so that they can be tracked and their potential impacts accounted for in applications where appropriate.
Each item will be reviewed as part of each STI change assessment that is performed and an assessment of the impact on the results of the application will be made prior to presenting the results of the risk analysis to the IDP. If a non-trivial impact is expected, then this may include the performance of additional sensitivity studies or PRA model changes to confirm the impact on the risk analysis.
2.0.5 Identification of Key Assumptions The overall Initiative 5b process is a risk-informed process with the PRA model results providing one of the inputs to the IDP to determine if an STI change is warranted. The methodology recognizes that a key area of uncertainty for this application is the standby failure rate utilized in the determination of the STI extension impact. Therefore, the methodology requires the performance of selected sensitivity studies on the standby failure rate of the component(s) of interest for the STI assessment.
The results of the standby failure rate sensitivity study plus the results of any additional sensitivity studies identified during the performance of the reviews as outlined in 2.2.1 and 2.2.3 above for each STI change assessment will be documented and included in the results of the risk analysis that goes to the IDP.
2.1 External Events Considerations The NEI 04-10 [Ref. 1] methodology allows for STI change evaluations to be performed in the absence of quantifiable PRA models for all external hazards. For those cases where the STI
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 6 of 29 cannot be modeled in the plant PRA (or where a particular PRA model does not exist for a given hazard group), a qualitative or bounding analysis is performed to provide justification for the acceptability of the proposed test interval change.
External hazards were evaluated in the GINNA Individual Plant Examination of External Events (IPEEE) submittal in response to the NRC IPEEE Program (Generic Letter 88-20, Supplement
- 4) [Ref. 41. The IPEEE Program was a one-time review of external hazard risk and was limited in its purpose to the identification of potential plant vulnerabilities and the understanding of associated severe accident risks.
The primary areas of external event evaluation at GINNA were internal fires and seismic risk.
The internal fire events were addressed by using a combination of the EPRI Fire Induced Vulnerability Evaluation (FIVE) methodology [Ref. 141 and fire PSA. The results of the Fire Analysis are documented in the R.E. Ginna Nuclear Power Plant IPEEE Fire Analysis transmitted to the NRC in June 1998 [Ref. 131. The seismic evaluations were performed in accordance with Generic Implementation Procedure (GIP) developed by the Seismic Qualification Utility Group (SQUG) of which Ginna was a member. The GIP provided plants a method for addressing Unresolved Safety Issue A-46 (Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors (USI A-46). Beyond this, Ginna performed a reduced-scope IPEEE for seismic events to close out IPEEE for Seismic Events.
The Ginna USI A-46 Seismic Evaluation Report and the IPEEE Seismic Evaluation Report were transmitted to the NRC in January 1997 [Ref. 151 However, there are no comprehensive CDF and LERF values available from the seismic IPEEE report to support the STI risk assessments.
High Winds, External Floods and Transportation Accidents were reviewed against the Standard Review Plan (SRP) as Ginna was one of the eleven participants in the NRC's Systematic Evaluation Program (SEP). Following plant modifications, it was determined that the Ginna plant met the Standard Review Plan criteria. Based on the NRC Safety Evaluation Reports (SERs) for Ginna's SEP results, no further submittals for GL 88-20 Supplement 4 were warranted for high winds, external floods, or transportation accidents.
Since the performance of the IPEEE, Ginna has submitted a License Amendment Request for conversion from appendix R compliance to NFPA-805 for fire protection [Ref. 121. Pursuant to this change, a fire PRA has been created and implemented at GINNA. This Fire PRA model was created under the auspices of NUREG/CR-6850 [Ref. 161 and has undergone PWROG peer review (completed August 2012) [Ref. 111. The Ginna Fire PRA was developed using the National Institute of Standards and Technology (NIST) Consolidated Model of Fire and Smoke Transport (CFAST) Methodology [Ref. 181; the Fire Dynamics Simulator, also developed by NIST; NUREG-1805 Fire Dynamics Tools (FDTs) computational Spreadsheets [Ref. 21];
EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities [Ref. 161 and the associated NUREG/CR-6850 Frequently Asked Questions (FAQ) Process [Ref. 201; Fire Events Database [Ref. 191 and plant specific data. This fire PRA has numerous capabilities not considered in the IPEEE fire PRA model including explicit analysis of all risk significant fire areas such as the main control room (MCR) and Relay Room (RR). Multiple spurious operations (MSO) considerations are also included. The ignition frequencies for all fire areas were developed using the guidance in NUREG/CR-6850 [Ref. 16] and also incorporate revised guidance for ignition frequencies [Ref. 171.
As stated earlier, the NEI 04-10 [Ref. 1 methodology allows for STI change evaluations to be performed in the absence of quantifiable PRA models for all external hazards. Therefore, in performing the assessments for the other hazard groups, a qualitative or a bounding approach will be utilized in most cases. Where applicable, the results of any STI change will be evaluated
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 7 of 29 against this model to ensure there is no undue risk associated with a given STI change. This approach is consistent with the accepted NEI 04-10 methodology.
2.2 Summary The GINNA PRA technical capability evaluations and the maintenance and update processes described above provide a robust basis for concluding that the GINNA PRA is suitable for use in risk-informed processes such as that proposed for the implementation of a Surveillance Frequency Control Program. Also, in addition to the standard set of sensitivity studies required per the NEI 04-10 [Ref. 1 methodology, open items for changes at the site and remaining gaps to specific requirements in the PRA standard will be reviewed to determine which, if any, would merit application-specific sensitivity studies in the presentation of the application results.
2.3 References
[1] Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Industry Guidance Document, NEI 04-10, Revision 1, April 2007.
[2] Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, Revision 1, January 2007.
[3] American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-Sc-2007, New York, New York, July 2007.
[4] NRC Generic Letter 88-20, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Supplement 4," June 28, 1991.
[5] NEI 05-04, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard
[6] NEI 00-002 Probabilistic Risk Assessment (PRA) Peer Review Process Guidance, Revision 1, Nuclear Energy Institute (NEI), Washington, DC, May 2006
[7] RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements for R. E.
Ginna Station Probabilistic Risk Assessment, Project PA-RMSC-0386, August 2009.
[8] Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard for L1/LERF PRA for NPP Applications for the Ginna Station Fire PRA, Project PA-RMSC-0403, August 2012.
[9] NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, Revision 1, Nuclear Energy Institute (NEI), Washington, DC, June 2010.
[10] American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME PRA Standard (ASME/ANS RA-Sa-2009),
New York, New York, July 2009.
[11] Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard for Level l/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the Ginna Station Fire Probabilistic Assessment, August 2012.
[12] Letter from Mr. Joseph E. Pacher (Ginna LLC) to Document Control Desk (NRC), dated March 28, 2013, License Amendment Request Pursuant to 10 CFR 50.90: Adoption of NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (ADAMS Accession No. ML13093A064).
[13] Letter from Robert C. Mecredy (RG&E) to Guy S. Vissing (NRC), 1. Ginna Station Fire IPEEE, RE Ginna Nuclear Power Plant; 2 Hydrogen Storage Facility, dated June 30, 1998.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 8 of 29
[14] Fire-Induced Vulnerability Evaluation (FIVE) Methodology Plant Screening Guide, EPRI TR-100370, Electric Power Research Institute, Final Report, April 1992.
[15] Letter from RC Mecredy (RG&E) to Guy S. Vissing (NRC), Resolution of GL 87-02, Supplement 1 and GL 88-20, Supplements 4 and 5 (Seismic Event Only) RG&E Corp, R.E. Ginna Nuclear Power Plant, dated January 31, 1997.
[16] EPRI/NRC-RES, Fire PRA Methodology for Nuclear Power Facilities, EPRI 1011989, NUREG/CR-6850, Final Report, September 2005.
[17] Fire Probabilistic Risk Assessment Methods Enhancements Supplement 1 to NUREG/CR-6850 and EPRI 1011989, EPRI 1019259, Electric Power Research Institute, December 2009.
[18] National Institute of Standards and Technology's (NIST) Consolidated Model of Fire Growth and Smoke Transport (CFAST) Version (6) (Jones et al., 2004)
[19] NSAC/179L, Electric Power Research Institute, Fire Events Database for U.S. Nuclear Power Plants, Rev. 1, January, 1993.
[20] Letter from John A. Grobe (NRR) to Alexander Marion (NEI), dated June 1, 2009, Path Forward in Resolving Frequently Asked Questions Related to NUREG/CR-6850.
Accession No. ML090920045.
[21] NUREG-1 805, Supplement 1, Volumes 1 & 2, Fire Dynamics Tools (FDTs) Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program.
[22] Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, Revision 2, March, 2009
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 9 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IE-C12 COMPARE results and EXPLAIN Open F&O IE-CIO-01: The Ginna Initiating Event Documentation only: Provide This item is a documentation
[2005: differences in the initiating event Notebook (Gl-IE-0001, Rev. 1) Section 4.3 comparison of core damage results issue. No impact on TSTF-425 IE-CIO] analysis with generic data sources to provides a cross-reference between the based on generic data cross- analysis.
URE 845 provide a reasonableness check of the Ginna Initiating Events and the NRC Rates referenced in Table 4-7.
results. Initiating Events in table 4-7. Table 4-7 nof cross-reference includes columns for NUREG/CR-5750 Category and NP-2230 EPRI/NUREG/CR-3862 PWR Category.
Table 4-8 provides a cross-reference between Ginna and similar PWR plants (Point Beach, Prairie Island, and Kewaunee).
An explanation of differences in Initiating Events between Ginna and similar PWRs is contained in the PRA Quantification (QU)
Notebook (G1-QU-0001, Rev. 0) Table 4-5 "Comparison of Ginna Core Damage Results to Similar Plants'. However, no explanation of differences between plant-specific initiating events and generic initiating events was located in either the Initiating Event Notebook (Gl-IE-0001, Rev. 1) or QU Notebook (G1-QU-0001, Rev. 0).
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 10 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IE-C15 CHARACTERIZE the uncertainty in the Open F&O IE-C13-01: Gi-IE-000 , PRA Documentation only: Include error This item is a documentation
[2005: initiating event frequencies and INITIATING EVENT (IE) NOTEBOOK, Section factors and brief discussion about IE issue and IE frequency IE-C13] PROVIDE mean values for use in the 5 documents assumptions and sources of frequency uncertainty, distribution evaluation. Changes quantification of the PRA results. uncertainty. However, section 5 does not will not impact the TSTF-425 provide or reference the parametric analysis uncertainty initiating event data distribution. For example, the distribution for TIGRLOSP is identified in the CAFTA model, newauto_65a-w-FId.caf, has having an EF of 7.39. However, no documentation for the error factor could be found.
Therefore, this SR is not met.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 11 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 SC-A2 SPECIFY the plant parameters (e.g., highest Open F&O SC-A2-01: The definition of core We agree with the peer reviewers that Over the typical complete loss of node temperature, core collapsed liquid damage documented in the Ginna-AS- the approach taken in the Ginna PRA is decay heat removal timing level) and associated acceptance criteria success criteria, the delta time Notebook-Rev-1 Section 2.2 is consistent overly conservative and not consistent (e.g., temperature limit) to be used in with the examples of measures for core with the requirements of Category II. between core uncovery and CET determining core damage. Select these damage suitable for Capability Category I as The peer reviewers suggested using a temperatures reach 12007F for parameters such that determination of defined in NUREG/CR-4550. For Category II core exit temperature of 1200°F for 30 30 minutes or 1800° peak center core damage is as realistic as practical, in a manner - consistent with current best Ginna could use the code-predicted core minutes as the criterion for core line is fairly small. As such, the practice. DEFINE computer code-predicted exit temperature >1,200°F for 30 min using damage, but we would recommend timing benefit is not expected to acceptance criteria with sufficient margin PCTRAN (code with simplified core using either that criterion or a peak be large except for the fast on the code-calculated values to allow for modeling (PWR)). core node temperature of 18007F. moving events such as large limitations of the code, sophistication of break LOCAs. For these events, Based on a review of the PCTRAN the models, and uncertainties in the results, it is likely that the 18007F peak we use the UFSAR success results, in a manner consistent with the core temperature would be reached criteria. Although this is not requirements specified under HLR-SC-B.
earlier than the time at which the core expected to be a significant Examples of measures for core damage suitable for Capability Category Il/111,that exit temperature would be greater effect, we do remain a have been used in PRAs, include (a) than 1200°F for 30 minutes. conservative CAT I. Therefore, collapsed liquid level less than 1?3 core the model used for TSTS-425 height or code-predicted peak core analysis may be conservative.
temperature >2,500°F (BWR) (b) collapsed liquid level below top of active fuel for a prolonged period, or code-pre-dicted core peak node temperature >2,200'F using a code with detailed core modeling; or code-predicted core peak node temperature
>1,800°F using a code with simplified (e.g.,
single-node core model, lumped para-meter) core modeling; or code-predicted core exit temperature >1,200°F for 30 min using a code with simplified core modeling (PWR).
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 12 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 SC-A4 IDENTIFY mitigating systems that are Complete F&O SC-A4 Operator action Add RCHFDX1BAF to the Event Tree No impact to TSTF 425. Action shared between units, and the RCHFDX1BAF (operator fails to align BAF TIU, as appropriate, placed in Event Tree TIU logic manner in which the sharing is given 1 of 2 PORVs and no charging) is not and Finding addressed.
included in the fault tree model. It appears that this event should be added in Event experience a common initiating event Tree TIU Sequence 5 Failures under gate (e.g., LOOP). TLFB.
This is an omission in the model and may affect CDF and LERF.
SY-A10 INCORPORATE the effect of variable Complete SY-Al 1 Gate TLFBHRD1 input to gate Review the Bleed and Feed modeling No impact as the Finding has
[SY-A11 success criteria (i.e., success criteria TL_FB for failure of Bleed and Feed models that change as a function of plant success as requiring 1 SI pump and 1 PORV. to ensure the system failures been addressed and the logic has
-2005] status) into the system modeling. The logic does not include 75 gpm charging appropriately reflect the success been updated and documented Example causes of variable system flow which is noted in the Success Criteria criteria, in the Success Criteria Notebook.
success criteria are notebook as required to support single PORV (a) different accidentscenarios, success. This was confirmed through Different success criteria are required discussion with Ginna PRA personnel.
for some systems to mitigate different accident scenarios (e.g., the number of The omission of a needed mitigating system pumps required to operate in some for support of the Bleed and Feed function systems is dependent upon the may underestimate the importance of these modeled initiating event), sequences for applications.
(b) dependence on other components.
Success criteria for some systems are also dependent on the success of another component in the system (e.g.,
operation of additional pumps in some cooling water systems is required if noncritical loads are not isolated).
(c) time dependence. Success criteria for some systems are time-dependent (e.g., two pumps are required to provide the needed flow early following an accident initiator, but only one is required for mitigation later following the accident).
(d) sharing of a system between units.
Success criteria may be affected when both units are challenged by the same initiating event (e.g., LOOP).
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 13 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 SY-A14 When identifying the failures in SY-Al 1 Complete SY-A13 Inconsistencies existed in the Review the need to add the No impact to TSTF 425. The
[SY-A13 INCLUDE consideration of all failure system modeling of the city water system. unavailability event in the SAFW dependencies for SAFW have modes, consistent 2005] Where used to support the GE-Betz system, System. been updated in the Ginna PRA.
with available data and model level of detail, except where excluded using a basic event for unavailability of city water the criteria in due to grid LOOP was added (basic event SY-A 15. CDAACITYWATER). This same event was not For example, added to the city water modeling for (a) active component fails to start support of the SAFW system.
(b) active component fails to continue to run (c) failure ot a closed component to open (d) failure of a closed component to remain closed (e) failure of an open component to close (f) failure of an open component to remain open (g) active component spurious operation (h) plugging of an active or passive component (i) leakage of an active or passive component
- 6) rupture of an active or passive component (k)internal leakage of a component
(/) internal rupture of a component (in) failure to provide signal/operate (e.g., instrumentation)
(n) spurious signal/operation (o) pre-initiator human failure events (see SY-A 16)
(p) other failures of a component to perform its required function 62
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 14 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 SY-A19 In the systems model, INCLUDE out- Open SY-A18 Ginna PRA System Notebooks Determine if any maintenance If shadowed unavailability is in-of-service unavailability for provides a list of all the modeled T&M terms
[SY-A18 components in the system model, in Section 3.4.C. Section 2.9 of the practices are performed that result in fact significantly affecting the 20051 unless screened, in a manner notebooks provide discussion of procedures overlapping unavailability of multiple unavailability numbers, then this consistent with the actual practices and and testing that result in Unavailability. The systems. If it is determined that would conservatively affect TSTF-history of the review of these sections found no instances of simultaneous unavailability that can result simultaneous unavailability is possible, 425 analysis. The most plant for removing equipment from service. from planned activities. However, the PRA model these occurrences as a single significant unavailabilities are engineer noted in a discussion that some unavailability event in the PRA or related to MSPI related functions (a) INCLUDE systems are shadowed in planned (1) unavailability caused by testing maintenance. There is not a specific justify why the unavailability is treated which are less likely to include when a component or system train is discussion on plant maintenance practices as separate events and include this as a conservative data.
reconfigured from its required accident within the (a)(4) program that would result in mitigating position such that the potential source of model uncertainty.
planned unavailability of multiple systems component cannot function as required OOS (i.e., EDG outages combined with AFW Also, consider adding a specific (2) maintenance events at the train motor driven pump outages to lower total risk question to the system engineers' level when procedures require isolating as opposed to performing the work independently), or of planned activities questionnaire for each system to the entire train resulting in multiple components OOS that do determine if there are planned for maintenance not violate technical specifications (e.g., two evolutions that represent simultaneous (3) maintenance events at a sub-train AFW pumps in maintenance or an AFW and level (i.e., between tagout boundaries, SAFW pump in maintenance at the same unavailability of multiple SSCs.
such as a functional equipment group) time). If work is done in this manner, it may when directed by procedures be appropriate to account for the (b) Examples of out-of-service unavailability of both SSCs in a single term.
unavailability to be modeled are as Modeling of station maintenance practices follows: that result in planned maintenance evolutions with more than a single PRA component (1) train outages during a work window OOS (i.e., shadowing equipment outages) for preventive/corrective maintenance can help to minimize the number of random (2) a functional equipment group (FEG) failure sequences and ensure there is not removed from service for "double counting" of unavailability in the PRA.
preventive/corrective maintenance (3) a relief valve taken out of service
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 15 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 HR-G3 When estimating HEPs EVALUATE the Complete F&O HR-G3-01: Details regarding certain Issue: In item (d) of CC II, Ill, clarify No impact to TSTF 425. The HRAs impact of the following plant-specific elements of the analysis were lacking in the that 'clarity' refers to the meaning of have been reviewed to ensure and scenario-specific performance HRA Calculator for a sufficient number of the cues, etc. In item (g) of CC II, IlI, the needed parameters for the shaping factors: HFEs to judge that this requirement was not clarify that complexity refers to both evaluation have been populated.
(a) quality [type (classroom or met. Evidence that the relevant aspects determining the need for and CBDM is now used as a max simulator) and frequency] of the cited in the SR are addressed for each HFE is executing the required response. function of CBDT and HCR/ORE.
operator training or experience critical to assuring that an appropriate RCHFDMAKEUP as a specific (b) quality of the written procedures analysis has been performed. This is Resolution: Cat I, II, and III: (d) degree example has a timing basis from and administrative controls particularly important in the case of HRA, of clarity of the meaning of cues / Key Input 51. When the (c) availability of instrumentation for which the methods are less indications annunciator model is used, there needed to take corrective actions straightforward than they are for many (g) complexity of detection, diagnosis is a clear discussion as to the (d) degree of clarity of the other parts of the PRA. and decision-making, and executing applicability.
cues/indications the required response.
(e) human-machine interface (f) time available and time required to complete the response (g) complexity of the required response (h) environment (e.g., lighting, heat, radiation) under which the operator is working (i) accessibility of the equipment requiring manipulation (j) necessity, adequacy, and availability of special tools, parts, clothing, etc.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 16 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 HR-I1 DOCUMENT the human reliability Complete F&O HR-I1-01: The bulk of the Documentation only. Same issue as for No impact to TSTF 425. This item analysis in a manner that facilitates documentation for the HRA is provided in HR-G3. has been addressed. See HR-G3.
PRA applications, upgrades, and peer the HRA Calculator. There are numerous review. areas in which the documentation is incomplete. The documentation should include a fuller discussion of the cues, bases for timing, specific procedure steps, and other aspects that could affect the analyses.
QU-B5 Fault tree linking and some other Open F&O QU-B5-01: In Section 3.1 of the QU Documentation only: Provide a The circular logic process is self-modeling approaches may result in Notebook, a mention is made that circular discussion in the Quantification revealing when a support gate is circular logic that must be broken logic checks were performed on the Notebook Section 3.1 of the added to the tree the CAFTA before the model is solved. BREAK the integrated top logic model to ensure it did methodology used to address circular software identifies a circular logic circular logic appropriately. Guidance not exist. An example is listed, but there is logic, issue. The circular logic is broken for breaking logic loops is provided in no further discussion. System notebooks by inserting as much of the logic NUREG/CR-2728 [2-13]. When reviewed generally state in Section 3.3 what clip into the tree as possible.
resolving circular logic, DO NOT was done when circular logic was identified, Providing more examples of this introduce unnecessary conservatisms but no discussion of the methodology was in the documentation is not or non-conservatisms. provided nor how conservatisms or non- expected to affect the TSTF-425 conservatisms are avoided. No evidence evaluation.
that the required analysis was not performed.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 17 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 LE-C2 INCLUDE realistic treatment of Open F&O LE-C2a-01: It is conservative to NOT There are limited operator actions that There are limited operator
[2005: feasible operator actions following take credit for operator actions post core could influence LERF at Ginna, so the actions that could influence LERF LE-C2a] the onset of core damage consistent damage. This is a requirement of the effect of such actions is not likely to be at Ginna, so the effect of such with applicable procedures, e.g., standard to move from Category I to significant. Moreover, it is likely that actions is not likely to be EOPs/SAMGs, proceduralized actions, Category II. there will not be a need for a Category significant. If post-core-damage or Technical Support Center guidance. II rating in this area to meet the operator actions are credited, requirements for most risk-informed LERF estimates could be reduced, applications. One approach to but the impact would be reaching Category II would be to minimal. The omission of these include post-core damage operator operator actions is conservative actions in the PRA. It is also possible and does not adversely impact that simply identifying operator the use of the model for TSTF-actions and showing quantitatively 425 analysis.
that they will have a negligible impact on LERF will be sufficient to meet the requirements of Category II.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 18 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 LE-C11 JUSTIFY any credit given for N/A F&O LE-C9a-01: It does not appear that The requirement is to justify credit As no equipment or HRA is
[2005: equipment survivability or human credit was taken for continued operation of taken for equipment survivability or credited post-containment LE-C9a] actions that could be impacted by equipment and operator actions that could human actions that could be affected failure, the PRA model remains a containment failure, be impacted by containment failure. This is by containment failure. Since no such conservative CAT I.
a requirement of the standard to move credit was taken, the SR should have from Category I to Category I1. been judged as not applicable (N/A).
This is analogous to the assessment of LE-C7 (old LE-C6) which was judged by the peer reviewers as N/A because human actions that support the accident progression analysis were not credited. Also, note that, in the Calvert Cliffs peer review, the peer reviewers judged this SR as N/A for the same reason. Only if post-containment failure equipment operations or human actions are modeled in the future would it be necessary to provide engineering analysis and written justification as part of the PRA documentation. Otherwise, no additional work is needed.
LE-C13 PERFORM a containment bypass N/A F&O LE-CIO-01: Credit for scrubbing was Review the possible credit for release No impact to TSTF 425. A
[2005 analysis in a realistic manner. JUSTIFY not taken. A sensitivity for impact of scrubbing to reduce LERF. sensitivity for impact of LE-ClO] any credit taken for scrubbing (i.e., scrubbing was performed and it was scrubbing was performed and it provide an engineering basis for the determined that the impact of not was determined that the impact decontamination factor used). considering scrubbing is negligible. This is a of not considering scrubbing is requirement of the standard to move from negligible.
Category I to Category II.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 19 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 MU-D1 A PRA Configuration Control Program complete F&O MU-DI PRA Configuration Control The CRMP database has a placeholder This configuration control issue shall be in procedure (GNG-CM-1.01-3003) Step 5.13 for a listing of PRA applications. This has been addressed. No impact place. It shall contain the following provides guidance for updating risk- portion of the database has been to TSTF 425.
key elements: informed applications. The process populated to ensure all applications (a) a process for monitoring PRA described depends upon a database inputs and collecting requiring update following a model maintained by the Fleet PRA Services new information Supervisor to identify current living revision can be easily identified.
(b) a process that maintains and applications requiring change assessment upgrades the PRA other than those related to maintenance to be consistent with the as-built, as rule performance criteria. No such operated plant database could be identified for Ginna.
(c) a process that ensures that the cumulative impact of pending Without a current list of risk-informed changes is considered when applying applications, the maintenance and update the PRA process is dependent upon the knowledge (d) a process that maintains and experience of the staff to know which configuration control of computer applications require update. This creates codes used to support PRA the possibility that an application could be quantification missed in the update process.
(e) documentation of the Program
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 20 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFSO-A4 For each potential source of flooding complete F&O IF-B2-01: Failure mechanisms are Address the potential for human- No impact to TSTF 425.
[2005: water, IDENTIFY the flooding addressed in conjunction with the caused flooding in the Internal Discussion of human caused IF-82] mechanisms that would result in a calculation of flood frequencies, in Section Flooding Study (51 - 9100978 - 000). floods is discussed in detail in fluid release. INCLUDE: 5.2 of document 51-9100978-000. Failures Describe the situations where a human Section 3.3 and 5.3 of Internal (a) failure modes of components such of components in piping systems other than error could result in flooding (e.g., Flood Notebook (GI-IF-0000-rl) for thevarious analysessystems.
pro Based ed onon as pipes, tanks, gaskets, expansion tanks are explicitly addressed by the EPRI inadvertent valve opening, inadvertent the analyses performed, one joints, fittings, seals, etc. pipe failure data base. This was the source train realignment, doors left open) and maintenance induced flood was (b) human-induced mechanisms that employed to characterize the frequencies of estimate the probabilities of such added to the model, FL-ABO-M-could lead to overfilling tanks, floods for Ginna. There has, however, been events. Model such floods that cannot SW - 2,000 gpm SW flood in the diversion of flow through openings a very limited attempt to address human- be screened. Consistent with the Aux Building due to created to perform maintenance; induced flood mechanisms, as required by Standard, utilize generic data as maintenance, isolated within 65 inadvertent actuation of fire item (b) of SR IF-B2. required by SR IFEV-A7 (IF-D6 in 2005 minutes.
suppression system Standard)
(c) other events releasing water into Such events have been important causes of the area flooding in the operating experience for US nuclear power plants, and as noted above the assessment of such floods is explicitly required.
A more systematic consideration should be made of human-caused floods. This will need to include an assessment of generic data related to human-caused floods, per SR IF-D6.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 21 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFSN-A6 For the SSCs identified in iFSN-A5 Open F&O IF-C3-01: There is no discussion of Cat I1: INCLUDE failure by submergence Failures due to jet impingement
[2005: (2005 text: IF-C2c), IDENTIFY the failures due to jet impingement or pipe and spray in the identification process. and pipe whip are now discussed IF-C3] susceptibility of each SSC in a flood whip. There is limited consideration of ASSESS qualitatively the impact of in Section 3.3.1 of the Internal area to flood-induced failure failure due to humidity/high temperature flood-induced mechanisms that are Flood Notebook G2-1F-0000 ri.
mechanisms. due to failure of heating steam lines. There not formally Failures due to Spray are INCLUDE failure by submergence and is also no discussion of criteria employed to addressed (e.g., using the mechanisms discussed in Section 3.3.2.
spray in the identification process. consider the potential for spray failures. listed under Capability Category III of Impacts due to spray were EITHER: this requirement), by using assumed to exist within 10 feet a) ASSESS qualitatively the impact of To meet Capability Category II, it is conservative assumptions. of a break location. Spray events flood-induced mechanisms that are necessary either to provide at least a [SAIC note: these mechanisms include are discussed in the IF Flood not formally addressed (e.g., using qualitative assessment of the potential for submergence, spray, jet impingement, notebook Section 4.5. Two the mechanisms listed under jet impingement and pipe whip, or to state pipe whip, humidity, condensation, locations were identified in the Capability Category IIIof this that these failure mechanisms were not temperature concerns] Aux Building where Fire Service requirement), by using conservative considered. It is also required that potential Water could impact safety assumptions; OR spray failures be evaluated. While spray Revise the Internal Flooding Study (51 - related busses and these are b) NOTE that these mechanisms are failures are discussed, there are no criteria 9100978 - 000) to describe the criteria explicitly modeled (FL-ABM-FSW-not included in the scope of the specified that would provide assurance that used to determine the potential for BUS15 and FL-ABO-FSW-BUS14).
evaluation. they had been considered in a consistent failure resulting from spray. Reference URE 1179 documents that IF and adequately comprehensive manner. Appendix C for a listing of components Notebook needs Appendix C impacted by spray. Describe how completed to complete Provide the requisite discussion of pipe potential spray impact was addressed documentation of spray impacts whip and jet impingement to satisfy the in the model. Confirm that the and modeling of additional spray standard. Specify appropriate criteria for assignment of spray impact is floods if appropriate. This would spray impacts, and assure that the potential consistent with the criteria used. be evaluated for any potential spray failures adequately reflect these impacts to a surveillance criteria. In addition, include a qualitative frequency interval extension at discussion of the potential impact of the time of the evaluation but is jet impingement, pipe whip, humidity, not expected to have a condensation, and temperature significant impact.
effects.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 22 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFSN-A8 IDENTIFY inter-area propagation complete F&O IF-C3b-01: The analysis does not Cat II, I1l: IDENTIFY inter-area. A discussion of structural failure
[2005: through the normal flow path from document consideration of potential barrier of barriers credited as barriers IF-C3b] one area to another via drain lines; failures due to flooding loads (structural INCLUDE potential for structural failure has been added to the IF and areas connected via back flow failures, failures of doors, etc.) This is (e.g., of doors or walls) due to flooding Notebook rl, Section 4.2.1.
through drain lines involving failed required to meet capability categories loads and the potential for barrier check valves, pipe and cable beyond Capability Category I. unavailability, including maintenance penetrations (including cable trays), activities.
doors, stairwells, hatchways, and Review flood barriers and identify and HVAC ducts. INCLUDE potential for evaluate any whose failures could Include a discussion of the potential structural failure (e.g., of doors or contribute adversely to propagation of for barrier failure due to flooding, walls) due to flooding loads. flooding including structures and doors. For walls, a qualitative discussion would appear to be acceptable. For doors, however, specific failure criteria should be developed and described. Flood scenarios should be reviewed and revised, if necessary, to address the potential failure of doors.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 23 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFSN- USE potential human mitigative Open F&O IF-C8-01: Only one flood appears to Characterize in greater detail those The screened flood will be added A16 actions as additional criteria for have been screened based on qualitative potential human actions that could to the flood model (URE 1176).
[2005: screening out flood sources if all the consideration of potential human action; for terminate the event and develop an However, the impact is expected IF-C8] following can be shown: that action (2000 gpm FSW break in IBN), estimate of the likelihood of failing to to be minimal, and is not there doesn't appear to be any justification mitigate the pipe break using accepted expected to have any impact on (a) flood indication is available in the for the time identified (190 min). Nothing HRA methods. the SFCP.
control room; other than time available is cited as rationale for screening the event.
(b) the flood source can be isolated; and To meet Capability Category II, it is necessary to characterize potential human (c) the mitigative action can be actions that could terminate flooding more performed with high reliability for the explicitly than was done in this case.
worst flooding initiator (2005 text:
flood from that source). High Address the required aspects for this and reliability is established by any other human actions used in justifying demonstrating, for example, that the screening out flood scenarios.
actions are procedurally directed, that adequate time is available for response, that the area is accessible, and that there is sufficient manpower available to perform the actions.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 24 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFEV-A6 GATHER plant-specific information on Open F&O IF-D5a-01: The current analysis does Address potential issues with material Plant specific experience with
[2005: plant design, operating practices, and not adequately address plant-specific condition and water hammer using internal flooding, water hammer IF-D5a] conditions that may impact flood characteristics that might affect the manner plant-specific information. Use this is addressed in the IF Notebook likelihood (i.e., material condition of in which the frequencies of flooding are information to revise, if necessary, rev 1 in Sections 3.3. A fluid systems, experience with water estimated. piping failure frequencies available in discussion of Human-induced hammer, and maintenance-induced industry-wide sources, consistent with floods is contained in Section 5.3.
floods). In determining the flood- To meet Capability Category II, it is required the Standard. Regarding any effect on flood initiating event frequencies for flood that plant-specific information be collected frequency due to aging affects, a scenario groups, USE a combination and considered on a variety of aspects For maintenance-induced and other sensitivity evaluation for a of the following (2005 text does not (including material condition of fluid human-caused flooding, see IFSO-A4. particular STI evaluation would include "of the following") systems, experience with water hammer, URE 1153 was written to consider show if there was any impact.
and maintenance-induced floods). The updating flood frequency for aging (a) generic and plant-specific current analysis is limited to the use of affects based on EPRI-302000079 operating experience; generic failure rates. This is consistent with Rupture frequencies.
(b) pipe, component, and tank Capability Category I.
rupture failure rates from generic data sources and plant-specific Address potential issues with material experience; (2005 text: and) condition, experience with water hammer, (c) engineering judgment for etc. In particular, further attention should consideration of the plant-specific be paid to the possibility of maintenance-information collected. induced and other human-caused flooding.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 25 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFEV-A7 INCLUDE consideration of human- complete F&O IF-D6-01: Initiating events that could See IFSO-A4. Discussion of human caused
[2005: induced floods during maintenance result from human actions were considered floods is discussed in detail in IF-D6] through application of generic data. only for a small number of possible Section 3.3 and 5.3 of Internal maintenance activities. These flood Flood Notebook (G1-IF-0000-rl) contributors were not evaluated using for various systems. Based on generic data as required. the analyses performed, one maintenance induced flood was Operating experience for nuclear power added to the model, FL-ABO-M-plants has provided evidence that human- SW - 2,000 gpm SW flood in the caused floods can be important. The SR Aux Building due to requires that such floods be evaluated using maintenance, isolated within 65 at least generic data to meet Capability minutes.
Category I or II.
Perform a more detailed assessment of potential human-caused floods, and apply at least generic data to characterize their frequencies.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 26 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFEV-A8 SCREEN OUT flood scenario groups if complete F&O IF-D7-01: Quantitative screening of Update the Internal Flooding Study (51 No impact to TSTF 425. This issue
[2005: (a) the quantitative screening criteria some scenarios was performed, but it is not - 9100978 - 000) to describe the has been addressed. Internal IF-D7] in IFSN-A10 (2005 text: IE-C4), as clear what criteria were applied in doing so. criteria used to screen flood scenarios. Flood Notebook Section 4.6, applied to the flood scenario groups, The criteria should be defined and applied If current screening criteria are not Screening Scenarios and Sources, are met; OR in a clear and consistent manner. well defined, develop such criteria and was updated to document the (b) the internal flood-initiating event apply them to scenarios addressed in screening criteria used. Figure affects only components in a single SRs IF-D7 and IF-E3a provide explicit criteria the analysis. 4.1, was added which shows the system, AND it can be shown that the for performing quantitative screening of Screening Criteria and Table 4.6 product of the frequency of the flood flood scenarios. The IF Notebook was edited to show the screening and the probability of SSC failure documents that some scenarios were criterion used for various flood given the flood is two orders of screened on low frequency, but does not scenarios.
magnitude lower than: invoke any particular criteria in doing so.
the product of the non-flooding frequency for the corresponding Provide a clear set of criteria for performing initiating events in the PRA, AND the quantitative screening of flood scenarios, random (non-flood-induced) failure and apply the criteria in a clear and probability of the same SSCs that are consistent manner.
assumed failed by the flood.
If the flood impacts multiple systems, DO NOT screen on this basis.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 27 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 I FQU-A5 If additional human failure events are complete F&O IF-E5-01: It was not clear that the Re-examine each HFE included in the No impact to TSTF-425.
[2005: required to support quantification of requirements were met in all cases. For flooding analysis. Perform operator Ginna Station Flooding Human IF-ES] flood scenarios, PERFORM any human example, interviews to establish aspects interviews as needed or identify and Reliability Analysis (HRA) document previously performed interviews. documents the flood recovery reliability analysis in accordance with such as response times were apparently the applicable requirements performed as part of the flood analysis, but actions (Areva Document No.:
Required operator interviews should described in 2-2.5 (2005 text: Tables the HRA was dramatically changed and new 51-9099406-000 located in GSN comprise the following:
4.5.5-2(e) through 4.5.5-2(h)). interviews/changes were not incorporated, 1. evaluate the flooding events based on 0157). The information and HRA nor were any inputs obtained from the HRA similarities to identify aselect set of values in this notebook were performed as part of the flood analysis scenarios to review with the operators (for verified to be consistent with the carried forward. example, categorized by the system that HRA actions being used in the generated the flood, e.g., fire protection) internal flood model. No
- 2. schedule interview sessions of about 1/2 It is necessary to perform the assessment of additional interviews were hour to an hour per each flooding scenario, HFEs associated with internal flooding in the identified as being necessary.
conducted separately with two different same manner as for other HFEs. The operators (preferably one experienced, one requirements to confirm procedure paths, novice) to get diverse opinions.
timing, etc. via interviews with operators 3. include questions on timing consistent were not met for a number of events. with the HRA Calculator Time Window screen for time of cue, time to diagnosis, Re-examine the HFEs associated with time for execution/manipulation of action (including travel time, with potential flood-internal flooding, and either perform related access delays). Be sure to ask about needed operator interviews or identify and any differences for floods initiated in same document existing inputs. system but in different rooms.
- 4. document interviews during the sessions (notes and/or tape recordings) and later in the HRA Calculator screens for Operator Interviews and Time Window.
Estimate and document internal flooding HFEs using the same approach as was used for other HFEs in the PRA. Recalculate flood scenario frequencies based on the new HFEs.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 28 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFQU-B1 DOCUMENT the internal flood Open F&O IF-FI-O1: The documentation is comprised Documentation only: Revise the Internal This documentation item will not accident sequences and primarily of the internal flooding notebook, Flooding Study (51 - 9100978 - 000) to meet impact the TSTF 425 analysis.
[2005:
supplemented heavily with information provided the documentation requirements of the This item has largely been IF-F1] quantification in a manner that in a set of Excel worksheets. The notebook is 2009 Standard. Address NRC Resolutions as facilitates PRA applications, upgrades, addressed by adding tables in annotated to provide a link to elements of the appropriate.
and peer review. Section 5.2 that show the worksheets, and an "assumption" provides the It is recommended that the Study be development of each initiating formal tie between the notebook and the worksheets. Some areas in which the links were reformatted to be consistent with the HLRs event frequency, adding an indirect or missing were noted. and SRs of the Standard, integrating Initiating Event Summary Table appropriate parts of the worksheets into (section 5.2.17), adding a In general, the manner in which important parts the primary document. This will provide a simplified set of arrangement of the flood analysis are documented in what document that can be easily reviewed drawings showing each flood would usually be characterized as an informal set against the standard and easily followed by area (Appendix K), defining spray of worksheets is judged not to meet the personnel not involved in the original modeling criteria (Section 3.3.2) requirement that the analysis be documented in analysis.
and identifying for each flood a manner that facilitates applications, upgrades, and peer review. Consistent with the F&O, include the area whether it was screened following in the revised Study: and the screening criterion used In addition to developing a single integrated set - Include a set of simplified arrangement (Table 4.6). The remaining item of documentation for the internal flood analysis, drawings to explicate the definition of flood is to develop the criteria used to there were several areas in which additional areas and help in understanding aspects perform quantitative screening, documentation would make the analysis more such as flood propagation.
if applicable, in Section 6.0 (URE tractable have been provided in connection to - Tabulate the flood areas and identify 1177).
other SRs. These include the following: clearly which are screened and which
- Include a set of simplified arrangement retained for further analysis to make the drawings to explicate the definition of flood areas process more tractable. Specify clearly and help in understanding aspects such as flood which criteria (qualitative or quantitative) propagation. are employed in screening each flood area.
- Tabulate the flood areas and identify clearly - Define explicitly the criteria used to which are screened and which retained for perform quantitative screening as noted in further analysis to make the process more Section 6.0.
tractable. Specify clearly which criteria - Define the criteria used to determine (qualitative or quantitative) are employed in whether a PRA component was susceptible screening each flood area. to failure due to spray.
- Define explicitly the criteria used to perform quantitative screening as noted in Section 6.0.
- Define the criteria used to determine whether a PRA component was susceptible to failure due to spray.
LAR - Adoption of TSTF-425, Revision 3 Attachment 2 Docket No. 50-244 Page 29 of 29 Table 2-1 Internal Events PRA Peer Review - Findings SR Topic Status Finding/Observation Disposition Impact to TSTF-425 IFQU-B3 DOCUMENT sources of model Open F&O IF-F3-O1: Section 7 of the IF Notebook Documentation only: Update the This documentation issue will not
[2005: uncertainty and related assumptions provides a discussion of three areas discussion of assumptions and affect the TSTF 425 analysis. This IF-F3] (as identified in QU-E1 and QU-E2) considered to be major sources of uncertainty to be consistent with the issue has been partially associated with the internal flood uncertainty in the flood analysis. This does 2009 Standard. The 2005 Standard addressed by the calculation of accident sequences and not constitute an adequate characterization required the documentation of key error factors for the flood quantification. of the sources of uncertainty associated assumptions and key sources of initiating events. These have with the flood analysis or a comprehensive uncertainty, while the 2009 Standard been added to table 5-2, flood (2005 text: Document the key discussion of the assumptions that could eliminates the term "key." The frequencies. Remaining action is assumptions and the key sources of have an effect on the results. equivalent sections of other PRA to reference any key sources of uncertainty associated with the technical elements provide an example uncertainty from the EPRI internal flooding analysis.) A reasonably thorough investigation of of the detail that is required. In guideline on the treatment of sources of uncertainty is necessary for addition, the discussion of uncertainty uncertainty for ASME PRA proper characterization of the flood and impact of assumptions in the Standard SRs related to internal analyses and results. Quantification Notebook should be flooding in Section 7 of the PRA revised to include treatment of flood Internal Flooding Analysis System A more comprehensive characterization of issues (or alternately, a similar Notebook (URE 1178) sources of uncertainty, comparable to that treatment should be provided in the provided for other areas of the PRA, should Flood Notebook).
be developed for the internal flood analysis.
ATTACHMENT 3 License Amendment Request R. E. Ginna Nuclear Power Plant Docket No. 50-244 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
Pronosed Technical Snecification Paae Chanoes 1.1-4 3.3.6-2 3.5.1-2 3.7.12-1 3.1.1-1 3.4.1-1 3.5.2-2 3.7.14-1 3.1.2-2 3.4.1-2 3.5.2-3 3.8.1-3 3.1.4-3 3.4.2-1 3.5.4-1 3.8.1-4 3.1.5-1 3.4.3-2 3.6.2-4 3.8.1-5 3.1.6-2 3.4.4-1 3.6.3-6 3.8.3-2 3.1.8-2 3.4.5-2 3.6.3-7 3.8.4-2 3.2.1-3 3.4.6-2 3.6.4-1 3.8.6-1 3.2.1-4 3.4.7-2 3.6.5-1 3.8.6-2 3.2.2-2 3.4.8-2 3.6.6-2 3.8.7-2 3.2.3-1 3.4.9-1 3.6.6-3 3.8.8-2 3.2.4-3 3.4.11-3 3.7.2-2 3.8.9-2 3.3.1-8 3.4.12-4 3.7.4-1 3.8.10-2 3.3.1-9 3.4.12-5 3.7.5-3 3.9.1-1 3.3.1-10 3.4.13-2 3.7.6-1 3.9.2-2 3.3.2-3 3.4.14-2 3.7.7-2 3.9.3-2 3.3.2-4 3.4.14-3 3.7.8-2 3.9.4-2 3.3.3-2 3.4.15-3 3.7.9-2 3.9.5-2 3.3.4-2 3.4.16-2 3.7.10-1 3.9.6-1 3.3.5-3 3.5.1-1 3.7.11-1 5.5-13
Definitions 1.1 PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
- a. Described in Chapter 14, Initial Test Program of the UFSAR;
- b. Authorized under the provisions of 10 CFR 50.59; or
- c. Otherwise approved by the Nuclear Regulatory Commission (NRC).
PRESSURE AND The PTLR is the plant specific document that provides the reactor vessel TEMPERATURE pressure and temperature limits, including heatup and cooldown rates, LIMITS REPORT and the power operated relief valve lift settings and enable temperature (PTLR) associated with the Low Temperature Overpressurization Protection System for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6. Plant operation within these limits is addressed in individual specifications.
QUADRANT QPTR shall be the ratio of the highest average nuclear power in any POWER TILT quadrant to the average nuclear power in the four quadrants.
RATIO (QPTR)
RATED THERMAL RTP shall be a total reactor core heat transfer rate to the reactor coolant POWER of 1775 MWt.
(RTP)
SHUTDOWN SDM shall be the instantaneous amount of reactivity by which the reactor MARGIN is subcritical or would be subcritical from its present condition assuming:
(SDM)
- a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCAof highest reactivity worth, which is assumed to be fully withdrawn. With any RCCAs not capable of being fully inserted, the reactivity worth of the RCCAs must be accounted for in the determination of SDM; and
- b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal hot zero power temperature.
STAGGEREDT*EST A STAGGERED TEST BASIS shale eenii t ef the testing cf .n. .fthe BA&SI systems, subsystems,
.the. ehanncls, er designated ccmpononts during the finteryel speciflcd by the Gurveillenee Frogueney, se that all systems3, subsystems, e..the. designated.cmpncnt.
.hann.ls, arc tested during n Su,,,illaon F.equcney intervels, whcrc n is the tetal numbcr ef systems, subsystems, ehannela, er ether designated eenmpocnts gin the asseciated function.
R.E. Ginna Nuclear Power Plant 1.1-4 Amendment 100
SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)
LCO 3.1.1 SDM shall be within the limits specified in the COLR.
APPLICABILITY: MODE 2 with keff < 1.0, MODES 3, 4, and 5.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to restore 15 minutes SDM to within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM is within the limits specified in the COLR. 24 hot',cs INSERT 1 R.E. Ginna Nuclear Power Plant 3.1.1-1 Amendment 80
Core Reactivity 3.1.2 SURVEILLANCE FREQUENCY Y
SR 3.1.2.2 - NOTE -
- 1. Only required after 60 effective full power days (EFPD).
- 2. The predicted reactivity values must be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 EFPD after each fuel loading.
Verify measured core reactivity is within +/- 1 % Ak/k of 31 EFPP predicted values.
R.E. Ginna Nuclear Power Plant 3.1.2-2 Amendment 80
Rod Group Alignment Limits 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual rod positions within alignment limit. 1-2 heums SR 3.1.4.2 - NOTE -
Only required to be performed if the rod position deviation monitor is inoperable.
Verify individual rod positions within alignment limit. Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and eve~y hetUF -
theFeaftef A~- TI T11 SR 3.1.4.3 Verify rod freedom of movement (trippability) by 92-,ý moving each rod not fully inserted in the core to a MRPI transition in either direction.
SR 3.1.4.4 Verify rod drop time of each rod, from the fully Once prior to withdrawn position, is < 1.8 seconds from the reactor criticality beginning of decay of stationary gripper coil voltage to after each removal dashpot entry, with: of the reactor head
- a. Tavg _>500°F; and
- b. Both reactor coolant pumps operating.
R.E. Ginna Nuclear Power Plant 3.1.4-3 Amendment 80
Shutdown Bank Insertion Limit 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Shutdown Bank Insertion Limit LCO 3.1.5 The shutdown bank shall be at or above the insertion limit specified in the COLR.
- NOTE -
The shutdown bank may be outside the limit when required for performance of SR 3.1.4.3.
APPLICABILITY: MODE 1, MODE 2 with Keff > 1.0.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Shutdown bank not within A.1.1 Verify SDM is within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> limit. limits specified in the COLR.
OR A.1.2 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SDM to within limit.
AND A.2 Restore shutdown bank to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> within limit.
B. Required Action and B.1 Be in MODE 2 with Keff 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion < 1.0.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify the shutdown bank insertion is within the limit specified in the COLR. n INE R.E. Ginna Nuclear Power Plant 3.1.5-1 Amendment 80
Control Bank Insertion Limits 3.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify estimated critical control bank position is within Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior the limits specified in the COLR. to achieving criticality SR 3.1.6.2 Verify each control bank insertion is within the limits specified in the COLR.
SR 3.1.6.3 - NOTE -
Only required to be performed if the rod insertion limit monitor is inoperable.
Verify each control bank insertion is within the limits Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> specified in the COLR. and eveFy 44teF9-t-hefeefte lINSfERT 11 SR 3.1.6.4 Verify each control bank not fully withdrawn from the core is within the sequence and overlap limits specified in the COLR.
R.E. Ginna Nuclear Power Plant 3.1.6-2 Amendment 80
PHYSICS TESTS Exceptions - MODE 2 3.1.8 CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 15 minutes associated Completion Time of Condition C not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Perform a COT on power range and intermediate Once within 7 days range channels per SR 3.3.1.7 and SR 3.3.1.8. prior to criticality SR 3.1.8.2 Verify the RCS lowest loop average temperature is "3 !NStER
Ž5300 F. INSERT 1 SR 3.1.8.3 Verify THERMAL POWER is <5% RTP. 30-mifiutes i . 1.
SR 3.1.8.4 Verify SDM is within the limits specified in the COLR.
N-wTdINSERT1j R.E. Ginna Nuclear Power Plant 3.1.8-2 Amendment 80
FQ(Z) 3.2.1 SURVEILLANCE REQUIREMENTS
- NOTE -
During power escalation at the beginning of each cycle, THERMAL POWER may be increased I until an equilibrium power level has been achieved, at which a power distribution map is obtained.
SURVEILLANCE FREQUENCY I
SR 3.2.1.1 Verify FQc(Z) is within limit. Once after each refueling prior to THERMAL POWER exceeding 75%
RTP AND Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions after exceeding, by
>_10% RTP, the THERMAL POWER at which FQC(Z) was last verified AND I 31 E th1rIftc NSFPD R.E. Ginna Nuclear Power Plant 3.2.1-3 Amendment 94
FQ(Z) 3.2.1 SURVEILLANCE FREQUENCY
- NOTE -
SR 3.2.1.2 If measurements indicate that the maximum over z [FQC(Z) / K(Z)]
has increased since the previous evaluation of FQC(Z):
- a. Increase FQW(Z) by the greater of a factor of 1.02 or by an appropriate factor specified in the COLR and reverify FQW(Z) is within limits or
- b. Repeat SR 3.2.1.2 once per 7 EFPD until either
- a. above is met or two successive flux maps indicate that the maximum over z [FQC(Z) / K(Z) ]
has not increased.
Verify FQW(Z) is within limit. Once after each refueling prior to THERMAL POWER exceeding 75%
RTP AND Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions after exceeding, by
_>10% RTP, the THERMAL POWER at which FQW(Z) was last verified AND
... NSERT1 R.E. Ginna Nuclear Power Plant 3.2.1-4 Amendment 94
FNAH 3.2.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify FNAH is within limits specified in the COLR. Once after each refueling prior to THERMAL POWER exceeding 75%
RTP AND
- 34 *,-Il I IAfeaIteF SIR3.2.2.2 ------ --- NOE-----------------
Only required to be performed if one power range channel is inoperable with THERMAL POWER _>75%
RTP.
Verify FNAH is within limits specified in the COLR. Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and eefy-2.4
.. .. A ..
L4ýNSERT 1T1 R.E. Ginna Nuclear Power Plant 3.2.2-2 Amendment 80
AFD 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD)
LCO 3.2.3 The AFD in % flux difference units shall be maintained within the limits spcified in the COLR.
- NOTE -
The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.
I APPLICABILITY: MODE 1 with THERMAL POWER >_50% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits. A.1 Reduce THERMAL POWER 30 minutes to < 50% RTP.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD within limits for each OPERABLE excore 1- Iys channel. j INSE R.E. Ginna Nuclear Power Plant 3.2.3-1 Amendment 94
QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY II SR 3.2.4.1 - NOTE -
- 1. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER
< 75% RTP, the remaining three power range channels can be used for calculating QPTR.
- 2. SR 3.2.4.2 may be performed in lieu of this Surveillance. INSERT 1 Verify QPTR is within limit by calculation.
-- NOTE -
SR 3.2.4.2 Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER
> 75% RTP.
Perform SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.2.1.
R.E. Ginna Nuclear Power Plant 3.2.4-3 Amendment 94
RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME W.2 Restore trip mechanism or 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> train to OPERABLE status.
X. Required Action and X.1 Initiate action to fully insert Immediately associated Completion all rods.
Time of Condition W not met. AND X.2 Place the Control Rod Drive 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> System in a Condition incapable of rod withdrawal.
SURVEILLANCE REQUIREMENTS
- NOTE -
Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.
SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. 42- h,.....
SR 3.3.1.2 - NOTE -
Required to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after L-ýýý THERMAL POWER is Ž>50% RTP.
Compare results of calorimetric heat balance 24 heufs calculation to Nuclear Instrumentation System (NIS) channel output and adjust if calorimetric power is
> 2% higher than indicated NIS power.
4-SR 3.3.1.3 - NOTE -
- 1. Required to be performed within 7 days after THERMAL POWER is Ž 50% RTP but prior to exceeding 90% RTP following each refueling and if the Surveillance has not been performed within the last 31 EFPD.
- 2. Performance of SR 3.3.1.6 satisfies this SR.
INSERT 1]
Compare results of the incore detector measurements 31 effeet-vc ul to NIS AFD and adjust if absolute difference is > 3%. power days (EFPD9)
R.E. Ginna Nuclear Power Plant 3.3.1-8 Amendment 112
RTS Instrumentation 3.3.1 SURVEILLANCE FREQUENCY SR 3.3.1.4 Perform TADOT.
STACCERE ITEST BA I ' I ETT&
SR 3.3.1.5 Perform ACTUATION LOGIC TEST. ,4, days...a.
S- ACGEREDTEST BASERTS SR 3.3.1.6 - NOTE -
Not required to be performed until 7 days after THERMAL POWER is Ž_50% RTP, but prior to exceeding 90% RTP following each refueling. 2 INSERT 1 Calibrate excore channels to agree with incore 92-EFPPQ detector measurements.
SR 3.3.1.7 ------ NOTE -----------------
Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering MODE 3. INSERT 1 Perform COT. 92 days SR 3.3.1.8 ------ NOTE -----------------
- 1. Not required for power range and intermediate range instrumentation until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reducing power < 6% RTP.
- 2. Not required for source range instrumentation until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reducing power < 5E-11 amps. INSERT 1 Perform COT. 92 days
-- NOTE -
SR 3.3.1.9 Setpoint verification is not required. INSERT 1 Perform TADOT.
R.E. Ginna Nuclear Power Plant 3.3.1-9 Amendment 112
RTS Instrumentation 3.3.1 SURVEILLANCE FREQUENCY SR 3.3.1.10 - NOTE -
Neutron detectors are excluded.
Perform CHANNEL CALIBRATION. 24-Riei4he SR 3.3.1.11 Perform TADOT.
SR 3.3.1.12 - NOTE -
Setpoint verification is not required.
Perform TADOT. Prior to reactor startup if not performed within previous 31 days i
SR 3.3.1.13 Perform COT. OA -m-i6 INtSERT 1 R.E. Ginna Nuclear Power Plant 3.3.1-10 Amendment 112
ESFAS Instrumentation 3.3.2 CONDITION REQUIRED ACTION COMPLETION TIME L. As required by Required L.1 Action A. 1 and referenced ------------
by Table 3.3.2-1. - NOTE -
The inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of the other channels.
Place channel in trip. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> M. Required Action and M.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition L not AND met.
M.2 Reduce pressurizer 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> pressure to < 2000 psig.
N. As required by Required N.1 Declare associated Auxiliary Immediately Action A. 1 and referenced Feedwater pump inoperable by Table 3.3.2-1. and enter applicable condition(s) of LCO 3.7.5, "Auxiliary Feedwater (AFW)
System."
SURVEILLANCE REQUIREMENTS
- NOTE -
Refer to Table 3.3.2-1 to determine which SRs apply for each ESFAS Function.
SURVEILLANCE FREQUENCY SR 3.3.2.1 Perform CHANNEL CHECK. 4hf INSERT 1 SR 3.3.2.2 Perform COT. 92-~ys
- NOTE 4T SR 3.3.2.3 -
Verification of relay setpoints not required.
Perform TADOT.
L4INSERT 11 R.E. Ginna Nuclear Power Plant 3.3.2-3 Amendment 109
ESFAS Instrumentation 3.3.2 SURVEILLANCE FREQUENCY SR 3.3.2.4 - NOTE -
Verification of relay setpoints not required.
Perform TADOT. 24-mgenths SR 3.3.2.5 Perform CHANNEL CALIBRATION.
SR 3.3.2.6 Verify the Pressurizer Pressure-Low and Steam Line 24 Fflenths Pressure-Low Functions are not bypassed when pressurizer pressure > 2000 psig. t1E~Tl i
SR 3.3.2.7 Perform ACTUATION LOGIC TEST. OA --- 16e L-lJsERýT1lT R.E. Ginna Nuclear Power Plant 3.3.2-4 Amendment 109
PAM Instrumentation 3.3.3 CONDITION REQUIRED ACTION COMPLETION TIME D. One or more Functions D.1 Restore one channel to 7 days with two required OPERABLE status.
channels inoperable.
E. Required Action and E.1 Enter the Condition Immediately associated Completion referenced in Table 3.3.3-1 Time of Condition C or D for the channel.
not met.
F. As required by Required F.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action E. 1 and referenced in Table 3.3.3-1. AND F.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> G. As required by Required G.1 Initiate action to prepare Immediately Action E.1 and referenced and submit a special report.
in Table 3.3.3-1.
SURVEILLANCE REQUIREMENTS
- NOTE -
SR 3.3.3.1 and SR 3.3.3.2 apply to each PAM instrumentation Function in Table 3.3.3-1.
SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform CHANNEL CHECK for each required '34 instrumentation channel that is normally energized.
SR 3.3.3.2 Perform CHANNEL CALIBRATION.
R.E. Ginna Nuclear Power Plant 3.3.3-2 Amendment 90
LOP DG Start Instrumentation 3.3.4 SURVEILLANCE REQUIREMENTS
- NOTE -
When a channel is placed in an inoperable status solely for the performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provided the second channel maintains LOP DG start capability.
SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform TADOT. 31 days SR 3.3.4.2 Perform CHANNEL CALIBRATION with Limiting Safety System Settings (LSSS)(a) for each 480 V bus t SE T as follows: -- I
- a. Loss of voltage LSSS _>372.0 V and < 374.8 V with a time delay of Ž 2.13 seconds and
- 2.62 seconds.
- b. Degraded voltage LSSS __420.0 V and ___ 423.6 V with a time delay of> 68.1 seconds and < 125 seconds (@ 420 V) and _>71.8 seconds and
< 125 seconds (@ 423.6 V).
(a)
A channel is OPERABLE when both of the following conditions are met:
- 1. The absolute difference between the as-found Trip Setpoint (TSP) and the previous as-left TSP is within the CHANNEL CALIBRATION Acceptance Criteria. The CHANNEL CALIBRATION Acceptance Criteria is defined as:
las-found TSP - previous as-left TSPI < CHANNEL CALIBRATION uncertainty The CHANNEL CALIBRATION uncertainty shall not include the calibration tolerance.
- 2. The as-left TSP is within the established calibration tolerance band about the nominal TSP. The nominal TSP is the desired setting and shall not exceed the LSSS.
The LSSS and the established calibration tolerance band are defined in accordance with the Ginna Instrument Setpoint Methodology. The channel is considered operable even if the as-left TSP is non-conservative with respect to the LSSS provided that the as-left TSP is within the established calibration tolerance band.
R.E. Ginna Nuclear Power Plant 3.3.4-2 Amendment 109
Containment Ventilation Isolation Instrumentation 3.3.5 SURVEILLANCE REQUIREMENTS
- NOTE -
Refer to Table 3.3.5-1 to determine which SRs apply for each Containment Ventilation Isolation Function.
SURVEILLANCE FREQUENCY INSERT 1 f l SR 3.3.5.1 Perform CHANNEL CHECK. 24 heu~sf INSERT 1 SR 3.3.5.2 Perform COT. 92 days INSERT 1 SR 3.3.5.3 Perform ACTUATION LOGIC TEST. 24 ments i .......
INSERT 1 SR 3.3.5.4 Perform CHANNEL CALIBRATION. 24ffefits R.E. Ginna Nuclear Power Plant 3.3.5-3 Amendment 85
CREATS Actuation Instrumentation 3.3.6 CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Suspend movement of Immediately associated Completion irradiated fuel assemblies.
Time of Condition A or B not met during movement of irradiated fuel assemblies.
SURVEILLANCE REQUIREMENTS
- NOTE -
Refer to Table 3.3.6-1 to determine which SRs apply for each CREATS Actuation Function.
CHMIC11 I AK1111= CDCtHC
,.,J
- ,, v IL...,-,d,, . I "',',.,'>',--- __-lN~iL-r I 1 I SR 3.3.6.1 Perform CHANNEL CHECK. 12 k. ,, Iurs lNSERT 1-]
Zý,K J.J.0.z Perform CU I. ui- eays
-- NOTE -
SR 3.3.6.3 Verification of setpoint is not required. - NSERT 1
]
Perform TADOT. FetS SR 3.3.6.4 Perform CHANNEL CALIBRATION. 24-ment-I ~INET1 SR 3.3.6.5 Perform ACTUATION LOGIC TEST. 'A--e "
R.E. Ginna Nuclear Power Plant 3.3.6-2 Amendment 87
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEMS (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)
Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in the COLR.
- NOTE -
Pressurizer pressure limit does not apply during pressure transients due to:
- a. THERMAL POWER ramp > 5% RTP per minute; or
- b. THERMAL POWER step > 10% RTP.
APPLICABILITY: MODE 1.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> parameters not within parameter(s) to within limit.
limits.
B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is within limit specified in 12- ..U.S the COLR.
SR 3.4.1.2 Verify RCS average temperature is within limit 12 hO',rs specified in the COLR. jLINSET1 R.E. Ginna Nuclear Power Plant 3.4.1-1 Amendment 80
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE FREQUENCY
- NOTE -
SR 3.4.1.3 Required to be performed within 7 days after > 95%
RTP.
Verify RCS total flow rate is within the limit specified in 24 months the COLR.
R.E. Ginna Nuclear Power Plant 3.4.1-2 Amendment 80
RCS Minimum Temperature for Criticality 3.4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 RCS Minimum Temperature for Criticality LCO 3.4.22 Each RCS loop average temperature (Tavg) shall be > 540'F.
APPLICABILITY: MODE 1, MODE 2 with keff > 1.0.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Tavg in one or both RCS A.1 Be in MODE 2 with Keff 30 minutes loops not within limit. < 1.0.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify RCS Tavg in each loop > 540 0 F. Within 30 minutes prior to achieving criticality.
SR 3.4.2.2 - NOTE -
Only required if any RCS loop Tavg < 547 0 F and the low Tavg alarm is either inoperable or not reset.
Verify RCS Tavg in each loop > 540'F. Once within 30 minutes and thereaftef T;NERIýT R.E. Ginna Nuclear Power Plant 3.4.2-1 Amendment 80
RCS P/T Limits 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
- NOTE -
SR 3.4.3.1 Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and INSERT 1 hydrostatic testing.
Verify RCS pressure, RCS temperature, and RCS 39-if motes heatup and cooldown rates are within the limits specified in the PTLR.
R.E. Ginna Nuclear Power Plant 3.4.3-2 Amendment 80
RCS Loops - MODE 1 > 8.5% RTP 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Loops - MODE 1 > 8.5% RTP LCO 3.4.z Two RCS loops shall be OPERABLE and in operation.
APPLICABILITY: MODE 1 > 8.5% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of LCO not A.1 Be in MODE 1 < 8.5% RTP. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify each RCS loop is in operation. 42--IhNSERT F9 HT1 R.E. Ginna Nuclear Power Plant 3.4.4-1 Amendment 80
RCS Loops - MODES 1 < 8.5% RTP, 2, and 3 3.4.5 CONDITION REQUIRED ACTION COMPLETION TIME C. Both RCS loops C.1 De-energize all CRDMs. Immediately inoperable.
AND OR C.2 Suspend operations that Immediately No RCS loop in operation, would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1.
AND C.3 Initiate action to restore one Immediately RCS loop to OPERABLE status and operation.
JINSERT 1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loop is in operation. 12 heWFS SR 3.4.5.2 Verify steam generator secondary side water levels 42-.
are _>16% for two RCS loops. INT1 SR 3.4.5.3 Verify correct breaker alignment and indicated power are available to the required RCP that is not in operation.
R.E. Ginna Nuclear Power Plant 3.4.5-2 Amendment 112
RCS Loops - MODE 4 3.4.6 CONDITION REQUIRED ACTION COMPLETION TIME B. One RHR loop - NOTE -
inoperable. Required Action B.1 is not AND applicable if all RCS and RHR loops are inoperable Two RCS loops - and
- Condition
- - - - C is-entered.
inoperable. B.1 Be in MODE 5. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C. All RCS and RHR loops C.1 Suspend operations that Immediately inoperable, would cause introduction of coolant into the RCS with OR boron concentration less than required to meet the No RCS or RHR loop in SDM of LCO 3.1.1.
operation.
AND C.2 Initiate action to restore one Immediately loop to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS INSERT 1 J-1 SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RHR or RCS loop is in operation. 12 h4etes SR 3.4.6.2 Verify SG secondary side water level is _>16% for each required RCS loop. IN SR 3.4.6.3 Verify correct breaker alignment and indicated power days are available to the required pump that is not in AT-INSERT1 operation.
R.E. Ginna Nuclear Power Plant 3.4.6-2 Amendment 112
RCS Loops - MODE 5, Loops Filled 3.4.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR loop A.1 Initiate action to restore a Immediately inoperable, second RHR loop to OPERABLE status.
AND OR Both SGs secondary side water levels not within A.2 Initiate action to restore Immediately limits, required SG secondary side water levels to within limits.
B. Both RHR loops B.1 Suspend operations that Immediately inoperable, would cause introduction of coolant into the RCS with OR boron concentration less than required to meet the No RHR loop in SDM of LCO 3.1.1.
operation.
AND B.2 Initiate action to restore one Immediately RHR loop to OPERABLE status and operation.
JINSERT 1 SURVEILLANCE REQUIREMENTS SURVEILLANCE REQUENCY SR 3.4.7.1 Verify one RHR loop is in operation. 42--i'us SR 3.4.7.2 Verify SG secondary side water level is > 16% in the 1- -" ....
required SG. TfINSERT1 SR 3.4.7.3 Verify correct breaker alignment and indicated power :-days are available to the required RHR pump that is not in /LINET operation.
R.E. Ginna Nuclear Power Plant 3.4.7-2 Amendment 112
RCS Loops - MODE 5, Loops Not Filled 3.4.8 CONDITION REQUIRED ACTION COMPLETION TIME B.2 Initiate action to restore one Immediately RHR loop to OPERABLE status and operation.
JINSERT 1 SURVEILLANCE REQUIREMENTS SURVEILLANCE REQUENCY SR 3.4.8.1 Verify one RHR loop is in operation. 12 h9'8r, SR 3.4.8.2 Verify correct breaker alignment and indicated power 7- dys are available to the RHR pump that is not in/LINSERT 1 operation.
R.E. Ginna Nuclear Power Plant 3.4.8-2 Amendment 112
Pressurizer 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LCO 3.4.9 The pressurizer shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer water level A.1 Be in MODE 3 with reactor 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> not within limit, trip breakers open.
AND A.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B. Pressurizer heaters B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> capacity not within limits.
AND B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS R.E. Ginna Nuclear Power Plant 3.4.9-1 Amendment 80
Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 - NOTE -
Not required to be performed with block valve closed per LCO 3.4.13.
Perform a complete cycle of each block valve. 92 days SR 3.4.11.2 Perform a complete cycle of each PORV. 24 Fa nths L-ý R.E. Ginna Nuclear Power Plant 3.4.11-3 Amendment 88
LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 - NOTE -
Only required to be performed when complying with LCO 3.4.12.a.
Verify no SI pump is capable of injecting into the RCS. k* INSERT 1 SR 3.4.12.2 - NOTE -
Only required to be performed when complying with LCO 3.4.12.b.
Verify a maximum of one SI pump is capable of 12-heuF s injecting into the RCS.
SR 3.4.12.3 - NOTE -
Only required to be performed when ECCS accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR.
Verify each ECCS accumulator motor operated Once within 12 isolation valve is closed. hours and evey- 12 SR 3.4.12.4 - NOTE -
Only required to be performed when complying with ;-eI'NSEF T 1 LCO 3.4.12.b.
Verify RCS vent > 1.1 square inches open. 12 he..s-for unlocked open vent valve(s)
- for locked open vent valve(s)
SR 3.4.12.5 Verify PORV block valve is open for each required PORV. NT R.E. Ginna Nuclear Power Plant 3.4.12-4 Amendment 88
LTOP System 3.4.12 SURVEILLANCE FREQUENCY
- NOTE -
SR 3.4.12.6 Required to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after decreasing RCS cold leg temperature to less than or equal to the LTOP enable temperature specified in the IST PTLR.
Perform a COT on each required PORV, excluding 31-days actuation.
SR 3.4.12.7 - NOTE -
Only required to be performed when ECCS accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR.
Verify power is removed from each ECCS Once within 12 accumulator motor operated isolation valve operator. hours and eyeiy 31 daysER 1h~e IN T SR 3.4.12.8 Perform CHANNEL CALIBRATION for each required 24 Ffents PORV actuation channel.
tf7J-T R.E. Ginna Nuclear Power Plant 3.4.12-5 Amendment 88
RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 - NOTE -
- 1. Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
- 2. Not applicable to primary to secondary LEAKAGE.
Verify RCS operational LEAKAGE is within limits by 7-2 heWF I performance of RCS water inventory balance.
SR 3.4.13.2 -- -
- NOTE -
Not required to be performed Until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
Verify primary to secondary LEAKAGE is < 150 72 hUFS, I gallons per day through any one SG.
R.E. Ginna Nuclear Power Plant 3.4.13-2 Amendment 100
RCS PIV Leakage 3.4.14 CONDITION REQUIRED ACTION COMPLETION TIME A.2 Isolate the high pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 - NOTE-
- 1. Not required to be performed until prior to entering MODE 2 from MODE 3.
- 2. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided. INSERT 1 Verify leakage from each SI cold leg injection line and 24 m,,nths each RHR RCS PIV is equivalent to < 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm AND at an RCS pressure > 2215 psig and < 2255 psig.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action, flow through the valve, or maintenance on the valve R.E. Ginna Nuclear Power Plant 3.4.14-2 Amendment 80
RCS PIV Leakage 3.4.14 SURVEILLANCE FREQUENCY SR 3.4.14.2 - NOTE -
- 1. Not required to be performed until prior to entering MODE 2 from MODE 3.
- 2. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.
A 11 ..
Verify leakage from each SI hot leg injection line RCS PIV is equivalent to < 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS AND pressure > 2215 psig and < 2255 psig.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve actuation due to automatic or manual action, flow through the valve, or maintenance on the valve R.E. Ginna Nuclear Power Plant 3.4.14-3 Amendment 80
RCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS JINSRT1i -
SURVEILLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of containment ,2 h-ews atmosphere radioactivity monitors. FINET1 SR 3.4.15.2 Perform COT of containment atmosphere radioactivity 92-days monitors. I 1 SR 3.4.15.3 Perform CHANNEL CALIBRATION of the required 24 FAIII containment sump monitor. INSERT 1 SR 3.4.15.4 Perform CHANNEL CALIBRATION of containment 24-menths atmosphere radioactivity monitors.
R.E. Ginna Nuclear Power Plant 3.4.15-3 Amendment 88
RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific activity < 10O/E S SdEyq pCi/gm.
SR 3.4.16.2 - NOTE -
Only required to be performed in MODE 1.
Verify reactor coolant DOSE EQUIVALENT 1-131 44-days specific activity < 1.0 pjCi/gm.
AND Between 2 and 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after a THERMAL POWER change of _>15%
RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period SR 3.4.16.3 - NOTE -
Only required to be performed in MODE 1.
Determine E from a reactor coolant sample. Once within 31 days after a minimum d 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
AND INSERT 1 rhe" eafte .
R.E. Ginna Nuclear Power Plant 3.4.16-2 Amendment 88
Accumulators 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators LCO 3.5.1 Two ECCS accumulators shall be OPERABLE.
APPLICABILITY: MODES 1 and 2, MODE 3 with pressurizer pressure > 1600 psig.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One accumulator A.1 Restore boron 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable due to boron concentration to within concentration not within limits.
limits.
B. One accumulator B.1 Restore accumulator to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> inoperable for reasons OPERABLE status.
other than Condition A.
C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B AND not met.
C.2 Reduce pressurizer 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> pressure to < 1600 psig.
D. Two accumulators D.1 Enter LCO 3.0.3. Immediately inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator motor operated isolation valve is fully open.
27E~qT1 SR 3.5.1.2 Verify borated water volume in each accumulator is 10 6--
_>1090 cubic feet (24%) and < 1140 cubic feet (83%).
r,---T
.41 lxi.-ý r- "
R.E. Ginna Nuclear Power Plant 3.5.1-1 Amendment 101
Accumulators 3.5.1 SURVEILLANCE SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is 12-heHFS
>_700 psig and < 790 psig.
ýER1
': ýINS
-I- ,F SR 3.5.1.4 Verify boron concentration in each accumulator is 12 heour (by I Ž_2550 ppm and < 3050 ppm. inleakage monitoring)
AND f-ienthe (by I sape INSERT 1 SR 3.5.1.5 Verify power is removed from each accumulator 31-days motor'operated isolation valve operator when pressurizer pressure is > 1600 psig.
R.E. Ginna Nuclear Power Plant 3.5.1-2 Amendment 101
ECCS - MODES 1, 2, and 3 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify the following valves are in the listed position. 4n --..--
Number Position Function 825A Open RWST Suction to SI Pumps 825B Open RWST Suction to SI Pumps 826A Closed BAST Suction to SI Pumps 826B Closed BAST Suction to SI Pumps 826C Closed BAST Suction to SI Pumps 826D Closed BAST Suction to SI Pumps 851A Open Sump B to RHR Pumps 851B Open Sump B to RHR Pumps 856 Open RWST Suction to RHR Pumps 878A Closed SI Injection to RCS Hot Leg 878B Open SI Injection to RCS Cold Leg 878C Closed SI Injection to RCS Hot Leg 878D Open SI Injection to RCS Cold Leg 896A Open RWST Suction to SI and Containment Spray 896B Open RWST Suction to SI and Containment Spray I ~I.
'4 SR 3.5.2.2 Verify each ECCS manual, power operated, and 81 days automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
F-EgKil SR 3.5.2.3 Verify each breaker or key switch, as applicable, for each MY-valve listed in SR 3.5.2.1, is in the correct position.
R.E. Ginna Nuclear Power Plant 3.5.2-2 Amendment 80
ECCS - MODES 1, 2, and 3 3.5.2 SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each ECCS pump's developed head at the test flow In accordance point is greater than or equal to the required developed with the head. Inservice Testing Program SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is nA ---
III LL-I ILI I not locked, sealed, or otherwise secured in position INSERT 1 actuates to the correct position on an actual or simulated actuation signal.
SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual 24 4:th or simulated actuation signal.
SR 3.5.2.7 Verify, by visual inspection, each RHR containment sump suction inlet is not restricted by debris and the 24-1T containment sump screen shows no evidence of structural distress or abnormal corrosion.
R.E. Ginna Nuclear Power Plant 3.5.2-3 Amendment 80
RWST 3.5.4 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.4 Refueling Water Storage Tank (RWST)
LCO 3.5.4 The RWST shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RWST boron A.1 Restore RWST to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> concentration not within OPERABLE status.
limits.
B. RWST water volume not B.1 Restore RWST to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> within limits. OPERABLE status.
C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS F INSERT 1 SURVEILLANCE FREQUENCY SR 3.5.4.1 Verify RWST borated water volume is > 300,000 7-days gallons (88%). INSERT 1I SR 3.5.4.2 Verify RWST boron concentration is > 2750 ppm and 7-days I < 3050 ppm.
R.E. Ginna Nuclear Power Plant 3.5.4-1 Amendment 96
Containment Air Locks 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1 - NOTE -
- 1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
- 2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.
Perform required air lock leakage rate testing in In accordance with accordance with the Containment Leakage Rate the Containment Testing Program. Leakage Rate Testing Program SR 3.6.2.2 Verify only one door in each air lock can be opened at 24 moneths a time.
R.E. Ginna Nuclear Power Plant 3.6.2-4 Amendment 80
Containment Isolation Boundaries 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each mini-purge valve is closed, except when 3j"eys the penetration flowpath(s) are permitted to be open NSERT1 under administrative control.
SR 3.6.3.2 - NOTE -
- 1. Isolation boundaries in high radiation areas may be verified by use of administrative controls.
- 2. Not applicable to containment isolation boundaries which receive an automatic containment isolation signal. INSERT I Verify each containment isolation boundary that is 92 days located outside containment and not locked, sealed, or otherwise secured in the required position is performing its containment isolation accident function except for containment isolation boundaries that are open under administrative controls.
- NOTE -
- 1. Isolation boundaries in high radiation areas may be verified by use of administrative means.
- 2. Not applicable to containment isolation boundaries which receive an automatic containment isolation signal.
Verify each containment isolation boundary that is Prior to entering located inside containment and not locked, sealed, or MODE 4 from otherwise secured in the required position is MODE 5 if not performing its containment isolation accident function, performed within the except for containment isolation boundaries that are previous 92 days open under administrative controls.
SR 3.6.3.4 Verify the isolation time of each automatic In accordance with containment isolation valve is within limits, the Inservice Testing Program SR 3.6.3.5 Perform required leakage rate testing of containment In accordance with mini-purge valves with resilient seals in accordance the Containment with the Containment Leakage Rate Testing Program. Leakage Rate Program.
R.E. Ginna Nuclear Power Plant 3.6.3-6 Amendment 80
Containment Isolation Boundaries 3.6.3 SURVEILLANCE FREQUENCY SR 3.6.3.6 Verify each automatic containment isolation valve that 24,,,,,st,,
is not locked, sealed, or otherwise secured in the required position actuates to the isolation position on "
an actual or simulated actuation signal.
R.E. Ginna Nuclear Power Plant 3.6.3-7 Amendment 80
Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCO 3.6.4t Containment pressure shall be > -2.0 psig and < 1.0 psig.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment pressure not A.1 Restore containment. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> within limits, pressure to within limits.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within limits.
TT1 N
R.E. Ginna Nuclear Power Plant 3.6.4-1 Amendment 80
Containment Air Temperature 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be < 125 0 F.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average air A.1 Restore containment 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> temperature not within average air temperature to limit. within limit.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is within 12 heH9ew limit.
R.E. Ginna Nuclear Power Plant 3.6.5-1 Amendment 116
CS, CRFC, and NaOH Systems 3.6.6 CONDITION REQUIRED ACTION COMPLETION TIME F. Two CS trains inoperable. F.1 Enter LCO 3.0.3. Immediately OR Three or more CRFC units inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Perform SR 3.5.2.1 and SR 3.5.2.3 for valves 896A In accordance with and 896B. applicable SRs.
SR 3.6.6.2 Verify each CS manual, power operated, and/,
- y automatic valve in the flow path that is not locked, INER sealed, or otherwise secured in position is in the correct position.
SR 3.6.6.3 Verify each NaOH System manual, power operated, 31-days and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the NSERLTl correct position.
SR 3.6.6.4 Operate each CRFC unit for _>15 minutes. 3 1 dIs- SERT 1 SR 3.6.6.5 Verify cooling water flow through each CRFC unit. 3t-y< SERTI SR 3.6.6.6 Verify each CS pump's developed head at the flow In accordance with test point is greater than or equal to the required the Inservice developed head. Testing Program SR 3.6.6.7 Verify NaOH System solution volume is _> 3000 gal. 84 4days ~ -- T SR 3.6.6.8 Verify NaOH System tank NaOH solution 1-84ýd*ys concentration is _> 30% and _<35% by weight. INSERT 1 SR 3.6.6.9 Perform required CRFC unit testing in accordance In accordance with with the VFTP. the VFTP SR 3.6.6.10 Verify each automatic CS valve in the flow path that is 24 menths not locked, sealed, or otherwise secured in position actuates to the correct position on an actual or simulated actuation signal.
R.E. Ginna Nuclear Power Plant 3.6.6-2 Amendment 99
CS, CRFC, and NaOH Systems 3.6.6 SURVEILLANCE FREQUENCY SR 3.6.6.11 Verify each CS pump starts automatically on an actual OA or simulated actuation signal. SERT1 SR 3.6.6.12 Verify each CRFC unit starts automatically on an actual or simulated actuation signal. 7A-- T1-SR 3.6.6.13 Verify each automatic NaOH System valve in the flow path that is not locked, sealed, or otherwise secured in position actuates to the correct position on an actual or simulated actuation signal.
SR 3.6.6.14 Verify spray additive flow through each eductor path. 6-yeeis SR 3.6.6.15 Verify each spray nozzle is unobstructed. Following maintenance which could result in nozzle blockage R.E. Ginna Nuclear Power Plant 3.6.6-3 Amendment 99
MSIVs and Non-Return Check Valves 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify closure time of each MSIV is < 5 seconds under In accordance with no flow and no load conditions, the Inservice Testing Program SR 3.7.2.2 Verify each main steam non-return check valve can In accordance with close, the Inservice Testing Program SR 3.7.2.3 Verify each MSIV can close on an actual or simulated 24 ,ionths actuation signal. INS.ERT R.E. Ginna Nuclear Power Plant 3.7.2-2 Amendment 80
ARVs 3.7.4 3.7 PLANT SYSTEMS 3.7.4 Atmospheric Relief Valves (ARVs)
LCO 3.7. I Two ARV lines shall be OPERABLE.
APPLICABILITY: MODES 1 and 2, MODE 3 with Reactor Coolant System average temperature (Tavg)
> 500 0 F.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I A. One ARV line inoperable. A.1 Restore ARV line to 7 days OPERABLE status.
B. Required Action and B.1 Be in MODE 3 with Tavg 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> associated Completion < 500 0 F.
Time of Condition A not met.
C. Two ARV lines C.1 Enter LCO 3.0.3. Immediately inoperable.
SURVEILLANCE REQUIREMENTS. IST SURVEILLANCE FREQUENCY SR 3.7.4.1 Perform a complete cycle of each ARV. 24 meths SR 3.7.4.2 Verify one complete cycle of each ARV block valve. P4 1NS;mei ET 1 L-ýNET 1 R.E. Ginna Nuclear Power Plant 3.7.4-1 Amendment 88
AFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Verify each AFW and SAFW manual, power operated, -ays 14 and automatic valve in each water flow path, and in both steam supply flow paths to the turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.
- NOTE -
SR 3.7.5.2 Required to be met prior to entering MODE 1 for the TDAFW pump.
Verify the developed head of each AFW pump at the In accordance with flow test point is greater than or equal to the required the Inservice developed head. Testing Program SR 3.7.5.3 Verify the developed head of each SAFW pump at the In accordance with flow test point is greater than or equal to the required the Inservice developed head. Testing Program SR 3.7.5.4 Perform a complete cycle of each AFW and SAFW In accordance with motor operated suction valve from the Service Water the Inservice System, each AFW and SAFW discharge motor Testing Program operated isolation valve, and each SAFW cross-tie motor operated valve.
SR 3.7.5.5 Verify each AFW automatic valve that is not locked, -,_4-e*_se4, sealed, or otherwise secured in position, actuates to INSERT 1 the correct position on an actual or simulated r actuation signal.
-- NOTE -
SR 3.7.5.6 Required to be met prior to entering MODE 1 for the TDAFW pump.
Verify each AFW pump starts automatically on an 24-ment actual or simulated actuation signal.
SR 3.7.5.7 Verify each SAFW train can be actuated and 244- negthe controlled from the control room. /I!NSERT" R.E. Ginna Nuclear Power Plant 3.7.5-3 Amendment 88
CSTs 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Condensate Storage Tanks (CSTs)
LCO 3.7.6 The CSTs shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. CST water volume not A.1 Verify by administrative 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> within limit, means OPERABILITY of backup water supply.
AND A.2 Restore CST water volume 7 days to within limit.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY I
SR 3.7.6.1 Verify the CST water volume is > 24,350 gal. 2 h...-...
- i==i si T 1 R.E. Ginna Nuclear Power Plant 3.7.6-1 Amendment 97
CCW System 3.7.7 CONDITION REQUIRED ACTION COMPLETION TIME D.2 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND D.3 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
- NOTE -
SR 3.7.7.1 Isolation of CtW flow to individual components does not render the CCW loop header inoperable.
Verify each CCW manual and power operated valve 31-days in the CCW train and heat exchanger flow path and loop header secured that is not in position, is inlocked, sealed, the correct or otherwise position. L NERT1 SR 3.7.7.2 Perform a complete cycle of each motor operated In accordance with isolation valve to the residual heat removal heat the Inservice exchangers. Testing Program R.E. Ginna Nuclear Power Plant 3.7.7-2 Amendment 80
SW System 3.7.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1 Verify screenhouse bay water level and temperature 24She R1s are within limits.
SR 3.7.8.2 ------ NOTE -----------------
Isolation of SW flow to individual components does not render the SW loop header inoperable.
Verify each SW manual, power operated, and 1-.. YS I automatic valve in the SW flow path and loop header L IERT1l that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.7.8.3 Verify all SW loop header cross-tie valves are locked , 1.days in the correct position. -- INSERT.l SR 3.7.8.4 Verify each SW automatic valve in the flow path that is ,4-met,*s not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual orINERT 1 simulated actuation signal.
SR 3.7.8.5 Verify each SW pump starts automatically on an actual or simulated actuation signal.
Z L. E1 R.E. Ginna Nuclear Power Plant 3.7.8-2 Amendment 102
CREATS 3.7.9 CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Place OPERABLE CREATS Immediately I associated Completion Time of Condition A not train in emergency mode.
met during movement of OR irradiated fuel assemblies. D.2 Suspend movement of Immediately irradiated fuel assemblies.
E. Two CREATS trains E.1 Suspend movement of Immediately inoperable during irradiated fuel assemblies.
movement of irradiated fuel assemblies.
OR One or more CREATS trains inoperable due to an inoperable CRE boundary during movement of irradiated fuel assemblies.
F. Two CREATS trains F.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2, 3, or 4 for reasons other than Condition B.
SURVEILLANCE REQUIREMENTS INSERT 1 SURVEILLANCE J FREQUENCY SR 3.7.9.1 Operate each CREATS filtration train > 15 minutes. ,4-,e..
SR 3.7.9.2 Perform required CREATS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.9.3 Verify each CREATS train actuates on an actual or 24-ment-hs simulated actuation signal. 'LIIN T1 R.E. Ginna Nuclear Power Plant 3.7.9-2 Amendment 105
ABVS 3.7.10 3.7 PLANT SYSTEMS 3.7.10 Auxiliary Building Ventilation System (ABVS)
LCO 3.7.10 The ABVS shall be OPERABLE and in operation.
APPLICABILITY: During movement of irradiated fuel assemblies in the Auxiliary Building when one or more fuel assemblies in the Auxiliary Building has decayed < 60 days since being irradiated.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ABVS inoperable. A.1
- NOTE -
LCO 3.0.3 is not applicable.
Suspend movement of Immediately irradiated fuel assemblies in the Auxiliary Building.
INSERT1 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.10.1 Verify ABVS is in operation. 24,*h,.FS SR 3.7.10.2 Verify ABVS maintains a negative pressure with 244r respect to the outside environment at the Auxiliary Building operating floor level.
SR 3.7.10.3 Perform required Spent Fuel Pool Charcoal Adsorber In accordance with System filter testing in accordance with the Ventilation the VFTP Filter Testing Program (VFTP).
R.E. Ginna Nuclear Power Plant 3.7.10-1 Amendment 80
SFP Water Level 3.7.11 3.7 PLANT SYSTEMS 3.7.11 Spent Fuel Pool (SFP) Water Level LCO 3.7.11 The SFP water level shall be >! 23 ft over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY: During movement of irradiated fuel assemblies in the SFP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SFP water level not within A.1 lim it.
- NOTE -
LCO 3.0.3 is not applicable.
Suspend movement of Immediately irradiated fuel assemblies in the SFP.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.11.1 Verify the SFP water level is Ž 23 ft above the top of 7-deys the irradiated fuel assemblies seated in the storage INET1 racks.
R.E. Ginna Nuclear Power Plant 3.7.11-1 Amendment 80
SFP Boron Concentration 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Spent Fuel Pool (SFP) Boron Concentration LCO 3.7.12 The SFP boron concentration shall be > 2300 ppm.
APPLICABILITY: Whenever any fuel assembly is stored in the SFP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SFP boron concentration -------------
- NOTE -
not within limit. LCO 3.0.3 is not applicable.
A. 1 Suspend movement of fuel Immediately assemblies in the SFP.
AND A.2 Initiate action to restore SFP Immediately boron concentration to within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Verify the SFP pool boron concentration is within limit. 7-deys R.E. Ginna Nuclear Power Plant 3.7.12-1 Amendment 80
Secondary Specific Activity 3.7.14 3.7 PLANT SYSTEMS 3.7.14 Secondary Specific Activity LCO 3.7.14 The specific activity of the secondary coolant shall be <ý 0. 10 PJCi/gm DOSE EQUIVALENT 1-131.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Specific activity not within A.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> limit.
AND A.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the specific activity of the secondary coolant is W-ye
- 0.10 pCi/gm DOSE EQUIVALENT 1-131. INSERT 1/
R.E. Ginna Nuclear Power Plant 3.7.14-1 Amendment 80
AC Sources - MODES 1, 2, 3, and 4 3.8.1 CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A, B, or AND C not met.
D.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> E. Two DGs inoperable. E.1 Enter LCO 3.0.3. Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power 7-deys availability for the offsite circuit to each of the 480 V LtINSERT 1 safeguards buses.
SR 3.8.1.2 - NOTE -
- 1. Performance of SR 3.8.1.9 satisfies this SR.
- 2. All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.
Verify each DG starts from standby conditions and achieves rated voltage and frequency.
R.E. Ginna Nuclear Power Plant 3.8.1-3 Amendment 109
AC Sources - MODES 1, 2, 3, and 4 3.8.1 SURVEILLANCE FREQUENCY SR 3.8.1.3 - NOTE -
- 1. DG loadings may include gradual loading as recommended by the manufacturer.
- 2. Momentary transients outside the load range do not invalidate this test.
- 3. This Surveillance shall be conducted on only one DG at a time.
- 4. This SR shall be preceded by and immediately follow without shutdown a successful performance of SR 3.8.1.2 or SR 3.8.1.9.
Verify each DG is synchronized and loaded and 31 days operates for _>60 minutes and < 120 minutes at a load I > 2025 kW and < 2250 kW. FýT~T SR 3.8.1.4 Verify the fuel oil level in each day tank. 31-days SR 3.8.1.5 Verify the DG fuel oil transfer system operates to 31days transfer fuel oil from each storage tank to the associated day tank.
SR 3.8.1.6 Verify transfer of AC power sources from the 50/50 mode to the 100/0 mode and 0/100 mode. SE
-- NOTE -
- 1. This Surveillance shall not be performed in MODE 1, 2, 3, or 4.
- 2. Credit may be taken for unplanned events that satisfy this SR. I Verify each DG does not trip during and following a 24-menths load rejection of Ž_295 kW.
R.E. Ginna Nuclear Power Plant 3.8.1-4 Amendment 109
AC Sources - MODES 1, 2, 3, and 4 3.8.1 SURVEILLANCE FREQUENCY i
SR 3.8.1.8 -NOTE-
- 1. This Surveillance shall not be performed in MODE 1, 2, 3, or 4.
- 2. Credit may be taken for unplanned events that satisfy this SR.
Verify each DG automatic trips are bypassed on an actual or simulated safety injection (SI) signal except:
17 Eý
- a. Engine overspeed;
- b. Low lube oil pressure; and
- c. Start failure (overcrank) relay.
SR 3.8.1.9 - NOTE -
- 1. All DG starts may be preceded by an engine prelube period.
- 2. This Surveillance shall not be performed in MODE 1, 2, 3, or 4.
- 3. Credit may be taken for unplanned events that satisfy this SR.
Verify on an actual or simulated loss of offsite power 24 menths signal in conjunction with an actual or simulated SI actuation signal:
- a. De-energization of 480 V safeguards buses;
- b. Load shedding from 480 V safeguards buses; and
- c. DG auto-starts from standby condition and:
- 1. energizes permanently connected loads,
- 2. energizes auto-connected emergency loads through the load sequencer, and
- 3. supplies permanently and auto-connected emergency loads for > 5 minutes.
R.E. Ginna Nuclear Power Plant 3.8.1-5 Amendment 109
Diesel Fuel Oil 3.8.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify each fuel oil storage tank contains > 5000 gal of diesel fuel oil for each required DG.
SR 3.8.3.2 Verify fuel oil properties of new and stored fuel oil are In accordance with tested in accordance with, and maintained within the the Diesel Fuel Oil limits of, the Diesel Fuel Oil Testing Program. Testing Program R.E. Ginna Nuclear Power Plant 3.8.3-2 Amendment 80
DC Sources - MODES 1, 2, 3, and 4 3.8.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is > 129 V on float charge. / ISR
- NOTE -
- 1. SR 3.8.4.3 may be performed in lieu of SR 3.8.4.2.
- 2. This Surveillance shall not be performed in MODE 1, 2, 3, or4.
Verify battery capacity is adequate to supply, and 24 fneoths maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test.
SIR 3.8.4.3 - NOTE -
This Surveillance shall not be performed in MODE 1, 2, 3, or4.
Verify battery capacity is > 80% of the manufacturer's 6E) n~enths rating when subjected to a performance discharge test. AND 12 months when battery shows degradation, or has reached 85% of expected life with capacity < 100% of manufacturer's rating AND 24 months when battery has reached 85% of the expected life with capacity _>100% of manufacturer's rating R.E. Ginna Nuclear Power Plant 3.8.4-2 Amendment 80
Battery Cell Parameters 3.8.6 3.8 ELECTRICAL POWER SYSTEMS 3.8.6 Battery Cell Parameters LCO 3.8.E5 Battery cell parameters for Train A and Train B batteries shall be within limits.
APPLICABILITY: MODES 1, 2, 3, and 4, When associated DC electrical power sources are required to be OPERABLE by LCO 3.8.5, "DC Sources - MODES 5 and 6."
ACTIONS
- NOTE -
Separate Condition entry is allowed for each battery.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more batteries A.1 Declare associated battery Immediately with one or more battery inoperable.
cell parameters not within limits.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify electrolyte level of each connected battery cell is above the top of the plates and not overflowing.
SR 3.8.6.2 Verify the float voltage of each connected battery cell 31-days is > 2.07 V. ---IINSET1 SR 3.8.6.3 Verify specific gravity of the designated pilot cell in 31-days each battery is >_1.195. 'I-IINSERT SR 3.8.6.4 Verify average electrolyte temperature of the 31-days designated pilot cell in each battery is _> 55 0 F. /'-IINSERT 1 SR 3.8.6.5 Verify average electrolyte temperature of every fifth cell of each battery is _>55 0 F. ' 5 INET1 R.E. Ginna Nuclear Power Plant 3.8.6-1 Amendment 80
Battery Cell Parameters 3.8.6 SURVEILLANCE FREQUENCY SR 3.8.6.6 Verify specific gravity of each connected battery cell 92-deys is: is:,'LIINSERT 11
- a. Not more than 0.020 below average of all connected cells, and
- b. Average of all connected cells is _>1.195.
R.E. Ginna Nuclear Power Plant 3.8.6-2 Amendment 80
AC Instrument Bus Sources - MODES 1, 2, 3, and 4 3.8.7 CONDITION REQUIRED ACTION COMPLETION TIME D. Two or more required D.1 Enter LCO 3.0.3. Immediately instrument bus sources inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct static switch alignment to Instrument Bus A and C.
SR 3.8.7.2 Verify correct Class 1E CVT alignment to Instrument Bus B. 'L fj jJ R.E. Ginna Nuclear Power Plant 3.8.7-2 Amendment 80
AC Instrument Bus Sources - MODES 5 and 6 3.8.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct static switch alignment to required AC 7-days instrument bus(es).
SR 3.8.8.2 Verify correct Class 1E CVT alignment to the required 7-0eys AC instrument bus. L-IY 1 R.E. Ginna Nuclear Power Plant 3.8.8-2 Amendment 112
Distribution Systems - MODES 1, 2, 3, and 4 3.8.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to 7 dpys required electrical power trains.T INSERT 1 R.E. Ginna Nuclear Power Plant 3.8.9-2 Amendment 80
Distribution Systems - MODES 5 and 6 3.8.10 CONDITION REQUIRED ACTION COMPLETION TIME A.2.1 Suspend CORE Immediately ALTERATIONS.
AND A.2.2 Suspend movement of IImmediately irradiated fuel assemblies.
AND A.2.3 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or boron concentration.
AND A.2.4 Initiate actions to restore Immediately required electrical power distribution train(s) to OPERABLE status.
AND A.2.5 Declare associated required Immediately residual heat removal loop(s) inoperable and not in operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to 7 days required electrical power distribution trains. !NSERT 1 R.E. Ginna Nuclear Power Plant 3.8.10-2 Amendment 112
Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of the Reactor Coolant System, the refueling canal, and the refueling cavity shall be maintained within the limit specified in the COLR.
APPLICABILITY: MODE 6.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration not A.1 Suspend CORE Immediately within limit. ALTERATIONS.
AND A.2 Suspend positive reactivity Immediately additions.
AND A.3 Initiate action to restore Immediately boron concentration to within limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is within the limit specified 721h-5s in the COLR. i INSERT 1 R.E. Ginna Nuclear Power Plant 3.9.1-1 Amendment 80
Nuclear Instrumentation 3.9.2 CONDITION REQUIRED ACTION COMPLETION TIME C.2 Suspend positive reactivity Immediately additions.
AND C.3 Perform SR 3.9.1.1 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perform CHANNEL CHECK. 42 heuFis SR 3.9.2.2 ---------------------------------- - --- INSERT 1
- NOTE -
Neutron detectors are excluded from CHANNEL CALIBRATION. kINSET1 Perform CHANNEL CALIBRATION. 24-me*the R.E. Ginna Nuclear Power Plant 3.9.2-2 Amendment 112
Containment Penetrations 3.9.3 CONDITION REQUIRED ACTION COMPLETION TIME A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify each required containment penetration is in the required status. Ž INERTIK SR 3.9.3.2 Verify each required containment purge and exhaust 24 meths valve actuates to the isolation position on an actual or /* NSERT 1 simulated actuation signal.
R.E. Ginna Nuclear Power Plant 3.9.3-2 Amendment 107
RHR and Coolant Circulation - Water Level Ž>23 Ft 3.9.4 CONDITION REQUIRED ACTION COMPLETION TIME A.4 Close all containment 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> penetrations providing direct access from containment atmosphere to outside atmosphere.
SURVEILLANCE REQUIREMENTS SURVEILLANCE { FREQUENCY SR 3.9.4.1 Verify one RHR loop is in operation and circulating 12 heUiF reactor coolant.
jNS-RT 1 R.E. Ginna Nuclear Power Plant 3.9.4-2 Amendment 112
RHR and Coolant Circulation - Water Level < 23 Ft 3.9.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 Verify one RHR loop is in operation and circulating 1" " .
reactor coolant. -'NSER T1 SR 3.9.5.2 Verify correct breaker alignment and indicated power l-ys available to the required RHR pump that is not in INSERT 1 operation.
R.E. Ginna Nuclear Power Plant 3.9.5-2 Amendment 112
Refueling Cavity Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Refueling Cavity Water Level LCO 3.9.( Refueling cavity water level shall be maintained _>23 ft above the top of reactor vessel flange.
APPLICABILITY: During movement of irradiated fuel assemblies within containment, During CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refueling cavity water A.1 Suspend CORE Immediately level not within limit. ALTERATIONS.
AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify refueling cavity water level is > 23 ft above the 24 hefurS top of reactor vessel flange. INSERT 1 R.E. Ginna Nuclear Power Plant 3.9.6-1 Amendment 80
Programs and Manuals 5.5
- e. The quantitative limits on unfiltered air inleakage into the CRE.
These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
- f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability and determining CRE unfiltered inleakage as required by paragraph c.
ýFý R.E. Ginna Nuclear Power Plant 5.5-13 Amendment 110
ATTACHMENT 4 License Amendment Request R. E. Ginna Nuclear Power Plant Docket No. 50-244 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
Proposed Technical Specification Bases Page Changes (NOTE: TS Bases pages below marked with an asterisk (*) do not contain any mark-ups.
These pages are provided for completeness and for information purposes only.)
B 3.1.1-5 B 3.3.1-46 B 3.4.9-4 B 3.6.2-7 B 3.7.14-3 B 3.9.6-3 B 3.1.2-5 B 3.3.1-47 B 3.4.11-7 B 3.6.3-11 B 3.8.1-12 B 3.1.4-8 B 3.3.2-31 B 3.4.12-10* B 3.6.3-12 B 3.8.1-13 B 3.1.4-9 B 3.3.2-32 B 3.4.12-11 B 3.6.3-14 B 3.8.1-14 B 3.1.5-5 B 3.3.2-33 B 3.4.12-12 B 3.6.4-3 B 3.8.1-15 B 3.1.6-6 B 3.3.2-34 B 3.4.12-13 B-3.6.5-3 B 3.8.1-16 B 3.1.8-7 B 3.3.3-16 B 3.4.13-4* B 3.6.6-8 B 3.8.3-3 B 3.1.8-8 B 3.3.3-17 B 3.4.13-5 B 3.6.6-9 B 3.8.4-6 B 3.2.1-9 B 3.3.4-7 B 3.4.13-6 B 3.6.6-10 B 3.8.4-7 B 3.2.1-10 B 3.3.5-8 B 3.4.14-5* B 3.6.6-11 B 3.8.4-8 B 3.2.1-11 B 3.3.5-9 B 3.4.14-6 B 3.7.2-6 B 3.8.6-3 B 3.2.2-5* B 3.3.6-7 B 3.4.14-7 B 3.7.4-4 B 3.8.6-4 B 3.2.2-6 B 3.3.6-8 B 3.4.15-5 B 3.7.5-8 B 3.8.7-6 B 3.2.3-3 B 3.4.1-4 B 3.4.15-6* B 3.7.5-9 B 3.8.8-5 B 3.2.4-5 B 3.4.1-5 B 3.4.16-4 B 3.7.5-10 B 3.8.9-9 B 3.2.4-6 B 3.4.2-3 B 3.4.16-5 B 3.7.6-3 B 3.8.10-6 B 3.3.1-39* B 3.4.3-6 B 3.5.1-6 B 3.7.7-6 B 3.9.1-4 B 3.3.1-40 B 3.4.4-3 B 3.5.1-7 B 3.7.8-7 B 3.9.2-3*
B 3.3.1-41 B 3.4.5-5 B 3.5.2-11 B 3.7.8-8 B 3.9.2-4 B 3.3.1-42 B 3.4.5-6 B 3.5.2-12 B 3.7.9-6 B 3.9.3-4 B 3.3.1-43 B 3.4.6-5 B 3.5.2-13 B 3.7.10-4 B 3.9.4-4 B 3.3.1-44 B 3.4.7-5 B 3.5.4-4 B 3.7.11-3 B 3.9.5-4 B 3.3.1-45 B 3.4.8-4 B 3.5.4-5 B 3.7.12-3 B 3.9.6-2*
SDM B 3.1.1 In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the flowpath of choice would utilize a highly concentrated solution, such as that normally found in the boric acid storage tank, or the refueling water storage tank. The operator should borate with the best source available for the plant conditions.
In determining the boration flow rate, the time in core life must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle when the boron concentration may approach or exceed 2000 ppm. Assuming that a value of 1% Ak/k must be recovered and a boration flow rate of 10 gpm using 13,000 ppm boric acid solution, it is possible to increase the boron concentration of the RCS by 100 ppm in approximately 35 minutes. If a boron worth of 10 pcm/ppm is assumed, this combination of parameters will increase the SDM by 1% Ak/k. These boration parameters of 10 gpm and 13,000 ppm represent typical values and are provided for the purpose of offering a specific example.
SURVEILLANCE SR 3.1.1.1 REQUIREMENTS In MODE 2 with Keff < 1.0 and MODES 3, 4, and 5, theSDM is verified by comparing the RCS boron concentration to a SHUTDOWN MARGIN requirement curve that was generated by taking into account estimated RCS boron concentrations, core power defect, control bank position, RCS average temperature, fuel burnup based on gross thermal energy generation, xenon concentration, samarium concentration, and isothermal temperature coefficient (ITC).
JINSERT 3 iD--ý in The rFrqu*n*y ef 24 heU.S is based en the gonor*ally slow *han,,
rcquirod beron eeneentratieon and the l6W probab3ility of E1n accidont
~curing ithout the Fcguircd 8DM. This allows time for the eporater to eollect thorcquircd data, which ineludcs peoferminig a boronl eemccntration analysis, and eemplcto the ealeuletien.
REFERENCES 1. Atomic Industrial Forum (AIF) GDC 27 and 28, Issued forcomment July 10, 1967.
- 2. "American National Standard Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," N18.2-1973.
- 3. UFSAR, Section 15.1.5.
- 4. UFSAR, Section 15.4.4.
R.E. Ginna Nuclear Power Plant B 3.1.1-5 Revision 42
Core Reactivity B 3.1.2 B._1 If the core reactivity cannot be restored to within the 1% Ak/k limit, or if the Required Actions of Condition A cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 with Keff < 1.0 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the SDM for MODE 2 with Keff < 1.0 is not met, then the boration required by SR 3.1.1.1 would occur. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 2 with Keff < 1.0 from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Core reactivity must be verified following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling). The comparison must be made prior to entering MODE 1 when the core conditions such as control rod position, moderator temperature, and samarium concentration are fixed or stable.
Since the reactor must be critical to verify core reactivity, it is acceptable to enter MODE 2 with Keff > 1.0 to perform this SR. This SR is modified by a Note to clarify that the SR does not need to be performed until prior to entering MODE 1.
SR 3.1.2.2 Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations. The comparison is made, considering that other core conditions are fixed or stable, including control rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. T:he-Frequemey of 31 EFPE), ia acceptable, based en the slow rate of core ehanges due to fuel depletion and the prcsenee ef ether indicat0rs (QPT-R, AF), et.) fr PrO.Mpt indic.tio f..an ano.maly. The SR is modified by two Notes. The first Note states that the SR is only required after 60 effective full power days (EFPD). The second Note indicates that the normalization of predicted core reactivity to the measured value must take place within the first 60 EFPD after each fuel loading. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations.
R.E. Ginna Nuclear Power Plant B 3.1.2-5 Revision 21
Rod Group Alignment Limits B 3.1.4 D.2 If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the plant conditions fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the plant must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the plant must be brought to at least MODE 2 with Keff < 1.0 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 2 with Keff < 1.0 from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.1.4.1 REQUIREMENTS VWeifiertion that individual rod poSitionS Wer wthin alignCint limits using MRPconr the PPCS at a F betweenty nf prvidcs a histFry that.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allews the operators to dtctrod a that is beginning t dcviatoe fremnits mpontod pesitien. This Frequency takes into account other rodpcsitien Doininformation that is available to the continuously eoratera in the eeotrol nrolom,that du a rirg atual rod m otion, d can imcandietly be eviations SR 3.1.4.2 moifiedT 3
When the rod position deviation monitor (i.e., the PPCS) is inoperable, no control room alarm imediatelyb is available between the normala12 het* Frequency to eece.IIISR alert the operators of a rod misalignment. A reduction of the Frequency te-4 heHF. provides sufficient monitoring of the rod positions when the monitor is inoperable. This Frequency takes into account other rod position information that is continuously available to the operator in the control room, so that during actual rod motion, deviations can immediately be detected.I- NSERT3 This SR is modified by a Note that states that performance of this SR is only necessary when the rod position deviation monitor is inoperable.
SR 3.1.4.3 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2with Keff Ž!1.0, tripping each control rod would result in radial or axial power tilts, or oscillations. Exercising each individual control rod every-92-days provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each control rod to a MRPI transition will not cause radial or axial power tilts, or oscillations, to occur. The 92 da'*I rEuRn3 y takes into considcratien ethe informatio R.E. Ginna Nuclear Power Plant B 3.1.4-8 Revision 60
Rod Group Alignment Limits B 3.1.4 available t3 the
-ep..t.. in the c.ntrol r...m and 3.1.4.1, "SR whihi pcrffrmed moroe frquontly and adds to the detcrmfinatien of OPERABIIT-*^ of thc rod During or between required performances of SR 3.1.4.3 (determination of control rod OPERABILITY by movement), if a control rod(s) is discovered to be immovable, but remains trippable and aligned, the control rod(s) is considered to be OPERABLE. At any time, if a control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken.
SR 3.1.4.4 Verification of rod drop times allows the operator to determine that the maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analysis. Measuring rod drop times prior to reactor criticality, after reactor vessel head removal, ensures that the reactor internals and rod drive mechanism will not interfere with rod motion or rod drop time, and that no degradation in these systems has occurred that would adversely affect control rod motion or drop time. This testing is performed with both RCPs operating and the average moderator temperature _>500OF to simulate a reactor trip under actual conditions.
This Surveillance is performed during a plant outage, due to the plant conditions needed to perform the SR and the potential for an unplanned plant transient if the Surveillance were performed with the reactor at power.
REFERENCES 1. Atomic Industrial Forum (AIF) GDC 6, 14, 27, and 28, Issued for comment July 10, 1967.
- 2. 10 CFR 50.46.
- 3. UFSAR, Chapter 15.
- 4. UFSAR, Section 15.4.6.
- 5. UFSAR, Section 15.1.5.
- 6. UFSAR, Section 15.4.2.
R.E. Ginna Nuclear Power Plant B 3.1.4-9 Revision 60
Shutdown Bank Insertion Limit B 3.1.5 plant to remain in an unacceptable condition for an extended period of time.
B._1 If Required Actions A.1 and A.2 cannot be completed within the associated Completion Times, the plant must be brought to a MODE where the LCO is not applicable. To achieve this status, the plant must be placed in MODE 2 with keff < 1.0 within a Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE REQUIREMENTS SR 3.1.5.1 JINSERT I-7 Since the shutdown bank is positioned manually by the control room operator, a verification of shutdown bank position at a Frequency ef-evety 12 he'rs is adequate to ensure that the bank is within the insertion limit.
Also, the 12 hIur F-rIquIc* y takes into i.c*u1nt IthlinlformtiIn available inthe eontrol rMOM for the purpese of moneitering the status of shutdeWn redS-.
REFERENCES 1. Atomic Industrial Forum (AIF) GDC 27, 28, 29, and 32, Issued for comment July 10, 1967.
- 2. 10 CFR 50.46.
- 3. UFSAR, Chapter 15.
- 4. UFSAR, Section 15.1.5.
- 5. UFSAR, Section 15.4.1.
- 6. UFSAR, Section 15.4.2.
- 7. UFSAR, Section 15.4.6.
R.E. Ginna Nuclear Power Plant B 3.1.5-5 Revision 60
Control Bank Insertion Limits B 3.1.6 SURVEILLANCE SR 3.1.6.1 REQUIREMENTS This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits. The Frequency of within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving criticality ensures that the estimated control bank position is within the limits specified in the COLR shortly before criticality is reached.
..ith an OPERABLE bank inscrtien limit molnitor (i.e., the control board annunciatorS, Yerification of the control bank insertion limits a Frcgueney of 12 hourS is sufflicint to cnSUrc OPERABILITY of the ben inscrtion limit moneiter and to detect eontrol banks that may be approaching the insertien limfits sinee, normally, Ycr; littlc red moltionl occur-s In 12 hourS.
SR 3.1.6.3 When the insertion limit monitor (i.e., the control board annunciatcrsl becomes inoperable, no control room alarm is available between the-norm.al 12 hOUr f" qu.n.y to alert the operators of a control bank not within the insertion limits. A reduction of the Frequency to every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provides sufficient monitoring of control rod insertion when the monitor is inoperable. Verification of the control bank position at a Frequency-heUFrS is sufficient to detect control banks that may be approaching the insertion limits. /tINSERT3] INSERT/ 1 This SR is modified by a Note that states that performance of this SR in only necessary when the rod insertion limit monitor is inoperable.
SR 3.1.6.4 When control banks are maintained within their insertion limits as required by SR 3.1.6.2 and SR 3.1.6.3 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the COLR. A Ir...u.ncy of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> .. is nsist.nt with the
^insertion limi-t hc*-k abov- in SR 3.1.6.2.
R.E. Ginna Nuclear Power Plant B 3.1.6-6 Revision 60
PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 D.1 If Required Action C.1 cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within an additional 15 minutes. The Completion Time of 15 additional minutes is reasonable, based on operating experience, for reaching MODE 3 from MODE 2 in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.1.8.1 REQUIREMENTS The power range and intermediate range neutron detectors must be verified to be OPERABLE in MODE 2 by LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." A CHANNEL OPERATIONAL TEST is performed on each power range and intermediate range channel within 7 days prior to criticality. This will ensure that the RTS is properly aligned to provide the required degree of core protection during the performance of the PHYSICS TESTS. The 7 day time limit is sufficient to ensure that the instrumentation is OPERABLE shortly before initiating PHYSICS TESTS.
SR 3.1.8.2 Verification that the RCS lowest loop Tavg is _>530'F will ensure that the plant is not operating in a condition that could invalidate the safety analyses. Control board indication for Tavg is available down to 5401F while indication from the plant process computer (PPCS) is available down to 5350 F. Between 530°F and 535 0 F, PPCS cold and hot leg indication should be used to determine Tavg.
Verifleaticn of the RCS tempefraturo at a Frcgucney ef 30 m~inutes durinlg the porfformanec of the PHYSICS TESTS will eMSurc that the nta eendkoecis ef the safety analyses arnt iatd SR 3.1.8.3 Verification that THERMAL POWER is:* 5% RTP using the NIS detectors will ensure that the plant is not operating in a condition that could invalidate the safety analyses. Verifeatien ef the THERMAL POWERa a Froqueney ef 30 mninutes during the peoreformanc ef the PHYSICS TESTS will onAuro that the minitial eenditieno of the safe" analyses are net 16 ele* d.e t- --- " .
R.E. Ginna Nuclear Power Plant B 3.1.8-7 Revision 34
PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 SR 3.1.8.4 The SDM is verified by comparing the RCS boron concentration to a SHUTDOWN MARGIN requirement curve that was generated by taking into account estimated RCS boron concentrations, core power defect, control bank position, RCS average temperature, fuel burnup based on gross thermal energy generation, xenon concentration, samarium concentration, and isothermal temperature coefficient (ITC).
The Frcequoney of 24 heurs is based en the gonorally slew chango in rcguirod beren eenecntraticn and en the low probability of an aee6idont
~curing ithout the roquwirod 69M.
REFERENCES 1. 10 CFR 50, Appendix B, Section XI.
- 2. 10 CFR 50.59.
- 3. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology Report," July 1985.
- 4. UFSAR, Section 14.6.
- 5. Letter from R. W. Kober (RGE) to T. E. Murley (NRC),
Subject:
"Startup Reports," dated July 9, 1984.
- 6. Letter from J. P. Durr (NRC) to B. A. Snow (RGE),
Subject:
"Inspection Report No. 50-244/88-06," dated April 28, 1988.
R.E. Ginna Nuclear Power Plant B 3.1.8-8 Revision 34
FQ(Z)
B 3.2.1 SR 3.2.1.1 Verification that FQC(Z) is within its specified limits involves increasing FQM(Z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain FQC(Z). Specifically, FQM(Z) is the measured value of FQ(Z) obtained from incore flux map results and FQC(Z) = FQM(Z) 1.0815 (Ref. 4). FQc(Z)is then compared to its specified limits.
The limit with which FQC(Z) is compared varies inversely with power above 50% RTP and directly with a function called K(Z) provided in the COLR.
Performing this Surveillance in MODE 1 prior to exceeding 75% RTP ensures that the FQC(Z)limit is met when RTP is achieved, because peaking factors generally decrease as power level is increased.
If THERMAL POWER has been increased by > 10% RTP since the last determination of FQc(Z), another evaluation of this factor is required 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions at this higher power level (to ensure that FQC(Z) values are being reduced sufficiently with power increase to stay within the LCO limits).
The Frcqu**iy .f 31 EFPD is adequate t, menitr the ,hangI ef pewre dliStributien with cerc burnup bcoausc such ehangco arc slew and we"l ecntrolled when the plant is eperotcd in aecordanee with the Tcchnical Spocifloations (TS).~
SR 3.2.1.2 t- ERT 3I The nuclear design process includes calculations performed to determine that the core can be operated within the FQ(Z) limits. Because flux maps are taken in steady state conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the flux map data. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation. The maximum peaking factor increase over steady state values, calculated as a function of core elevation, Z, is called W(Z). Multiplying the measured total peaking factor, FQC(Z), by W(Z) gives the maximum FQ(Z) calculated to occur in normal operation, FQW(Z).
R.E. Ginna Nuclear Power Plant B 3.2.1-9 Revision 42
FQ(Z)
B 3.2.1 The limit with which FQW(Z) is compared varies inversely with power above 50% RTP and directly with the function K(Z) provided in the COLR.
The W(Z) curve is provided in the COLR for discrete core elevations.
Flux map data are typically taken for 61 core elevations. FQW(Z) evaluations are not applicable for the following axial core regions, measured in percent of core height:
- a. Lower core region, from 0 to 8% inclusive and
- b. Upper core region, from 92 to 100% inclusive.
The top and bottom 8% of the core are excluded from the evaluation because of the low probability that these regions would be more limiting in the safety analyses and because of the difficulty of making a precise measurement in these regions.
This Surveillance has been modified by a Note that may require that more frequent surveillances be performed. If FQW(Z) is evaluated, an evaluation of the expression below is required to account fcr any increase to FQM(Z) that may occur and cause the FQ(Z) limit to be exceeded before the next required FQ(Z) evaluation.
If the two most recent FQ(Z) evaluations show an increase in the expression maximum over z [FQC(Z) / K(Z) ], it is required to meet the FQ(Z) limit with the last FQW(Z) increased by the greater of a factor of 1.02 or by an appropriate factor specified in the COLR or to evaluate FQ(Z) more frequently, each 7 EFPD. These alternative requirements prevent FQ(Z) from exceeding its limit for any significant period of time without detection.
Performing the Surveillance in MODE 1 prior to exceeding 75% RTP ensures that the FQ(Z) limit is met when RTP is achieved, because peaking factors are generally decreased as power level is increased.
FQ(Z) is verified at power levels >_10% RTP above the THERMAL POWER of its last verification, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions to ensure that FQ(Z) is within its limit at higher power levels.
The Gur.....ane. Fro.uen, y ef 31 EFPD is adequate to monitfr the ehange of pewer di~tributien with eero burnup. Thc Surveiilonoo may be dcne maer- frcguently if rcquircd by the rcoults ef F (Z)evluatie.
R.E. Ginna Nuclear Power Plant B 3.2.1-10 Revision 42
FQ(Z)
B 3.2.1 The Frogucncy ef 31 EFPID is adequate to moenitor the changc of pewcr diStribution beopuse sueh a ehango is sufflciently slew, when the plant is epcraitcd in eccordanco with the TS, to proolude adverse peaking faetefrs between 31 day supvcillanccs.
REFERENCES 1. 10 CFR 50.46.
- 2. UFSAR 15.4.5.4.3
- 3. Atomic Industrial Forum (AIF) GIDC-29, Issued for comment July 10, 1967
- 4. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.
- 5. WCAP-1 0216-P-A, Rev. 1lA, "Relaxation of Constant Axial Offset Control (and) FQ Surveillance Technical Specification," February 1994.
R.E. Ginna Nuclear Power Plant B 3.2.1-11 Revision 42
FNAH 3.2.2 A.3 Reduction in the Overpower AT and Overtemperature AT trip setpoints by
> 1% for each 1% by which FNAH exceeds its limit, ensures that continuing operation remains at an acceptable low power level with adequate DNBR margin. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is sufficient considering the small likelihood of a severe transient in this period, and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1.
A.4 Verification that FNAH has been restored within its limit by performing SR 3.2.2.1 or SR 3.2.2.2 prior to increasing THERMAL POWER above the limit imposed by Required Action A.1 ensures that the cause that led to the FNAH exceeding its limit is corrected, and core conditions during operation at higher power levels are consistent with safety analyses assumptions.
B. 1 If the Required Actions of A.1 through A.4 cannot be met within their associated Completion Times, the plant must be placed in a mode in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The value of FNAH is determined by using the movable incore detector system to obtain a flux distribution map. A data reduction computer program then calculates the maximum value of FNAH from the measured flux distributions. The measured value of FNAH must be multiplied by 1.04 to account for measurement uncertainty before making comparisons to the FNAH limit.
After each refueling, FNAH must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that FNAH limits are met at the beginning of each fuel cycle.
R.E. Ginna Nuclear Power Plant B 3.2.2-5 Revision 21
FNAH JINSERT 3 * /3.2.2 The Frequency of 31 EFPD) is acceptable becauise the power distribution changS relatively slowly ,ver thiS amouint of fuel burnup. J elII.lyl, AI this Frequency i. sho^t enough that th ,FNAH limnit .ann.t be e..eeded fr any signifi.ant pcri.
- d f operation. When the plant is already performing SR 3.2.2.2 to satisfy other requirements, SR 3.2.2.2 does not need to be suspended in order to perform SR 3.2.2.1 since the performance of SR 3.2.2.2 meets the requirements of SR 3.2.2.1.
SR 3.2.2.2 During power operation, the global power distribution is monitored by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," which are directly and continuously measured process variables.
With an NIS power range channel inoperable, QPTR monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased. Peforming GR 3.2.2.2 at a Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provides an accurate altornative mneans for eM5UFt~g-that-FNAH remains within limnits and the core power distribution is
.. ns stent with the safoty analyseos. A Frequoncy of 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,s ta Ikes into eansuderation the rate at which peaking factors are likely to ehange, an the time required......
to stabilize
....... the ,plant
,, and
, .... ,,...-.performf
,* ak flu1X MaP. +INSERT 3 i This Surveillance is modified by a Note, which states that it is required only when one power range channel is inoperable and the THERMAL POWER is >_75% RTP.
REFERENCES 1. 10 CFR 50.46.
- 2. UFSAR, Section 15.4.5.1.
- 4. American National Standard, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," N18.2-1973.
R.E. Ginna Nuclear Power Plant B 3.2.2-6 Revision 21
AFD B 3.2.3 APPLICABILITY The AFD requirements are applicable in MODE 1 greater than orequal to 50% RTP when the combination of THERMAL POWER and core peaking factors are of primary importance in safety analysis.
For AFD limits developed using RAOC methodology, the value of the AFD does not affect the limiting accident consequences with THERMAL POWER < 50% RTP and for lower operating power MODES.
ACTIONS A.1 As an alternative to restoring the AFD to within its specified limits, Required Action A.1 requires a THERMAL POWER reduction to < 50%
RTP. This places the core in a condtion for which the value of the AFD is not important in the applicable safety analyses. A Completion lime of 30 minutes is reasonable, based on operating experience, to reach 50%
RTP without challenging plant systems.
SURVEILLANCE SR 3.2.3.1 REQUIREMENTS This Surveillance verifies that the AFD, as indicated by the NIS excore channel, is within its specified limits. The *u.v.illan. F..quen.y ef 7 days is adequate eensidering that the AFD as menitorod by a comnputer and any deviatien frcmA roquircments iz Ealarmcfd.-
REFERENCES 1. WCAP-1 0216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control/FQ Surveillance Technical Specification", February 1994.
- 2. American National Standard, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," N18.2-1973.
- 3. UFSAR, Section 7.7.2.6.4.
R.E. Ginna Nuclear Power Plant B 3.2.3-3 Revision 42
QPTR B 3.2.4 assumptions, Required Action A.6 requires verification that FQ(Z) as approximated by FQC(Z) and FQW(Z), and FNAH are within their specified limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching equilibrium condition at RTP. As an added precaution, if the core power does not reach equilibrium condition at RTP within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, but it increases slowly, then the peaking factor surveillances must be performed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after increasing THERMAL POWER above the limit of Required Action A.1. These Completion Times are intended to allow adequate time to increase THERMAL POWER to above the limit of Required Action A.1, while not permitting the core to remain with unconfirmed power distributions for extended periods of time.
Required Action A.6 is modified by a Note that states that the peaking factor surveillances may only be done after the excore detectors have been normalized to eliminate the indicated tilt (i.e., Required Action A.5).
The intent of this Note is to have the peaking factor surveillances performed at operating power levels, which can only be accomplished after the excore detectors are adjusted to eliminate the indicated tilt and the core returned to power.
B.1 If Required Actions A.1 through A.6 are not completed within their associated Completion Times, the plant must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to < 50% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging plant systems.
SURVEILLANCE SR 3.2.4.1 REQUIREMENTS This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. TFIe-Frcqucncy of 7 days takes int3 acccun~t ethcr in9fefrmatien and 8olormFI available in the control room SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is
< 75% RTP and one power range channel is inoperable. Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1.
For those causes of quadrant power tilt that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of the core power tilt.
R.E. Ginna Nuclear Power Plant B 3.2.4-5 Revision 42
QPTR B 3.2.4 SR 3.2.4.2 This surveillance is modified by a Note, which states that it is not required until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the input from one or more Power Range Neutron Flux channel is inoperable and the THERMAL POWER is > 75% RTP.
With the input from a NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased.
When one NIS power range channel input is inoperable and THERMAL POWER is > 75% RTP, a full core flux map should be performed to verify the core power distribution instead ef using the thrc ,P-ERB- E poc, range chainncl inputs te Yerif; QPT-R by perfeFrming SR 3.2.1.1, SR 3.2.1.2 and SR 3.2.2.1, at a Fr un.y .. of 24 heur.. Performing a full core flux map provides an accurate alternative means for ensuring that FQ(Z) and FN AH remain within limits and the core power distribution is consistent with the safety analysis.
&t-TINET3 REFERENCES 1. 10 CFR 50.46.
- 2. UFSAR, Section 15.4.5.
- 3. Atomic Industrial Forum (AI) GDC 29, Issued for comment July 10,1967.
R.E. Ginna Nuclear Power Plant B 3.2.4-6 Revision 42
RTS Instrumentation B 3.3.1 X.1 and X.2 If the Required Action and Associated Completion Time of Condition W is not met, the plant must be placed in a MODE where the Functions are no longer required. To achieve this status, action be must initiated immediately to fully insert all rods and the CRD System must be incapable of rod withdrawal within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. These Completion Times are reasonable, based on operating experience to exit the MODE of Applicability in an orderly manner.
SURVEILLANCE The SRs for each RTS Function are identified by the SRs column of Tble REQUIREMENTS 3.3.1-1 for that Function.
A Note has been added to the SR Table stating that Table 3.3.1-1 determines which SRs apply to which RTS Functions.
Note that each channel of process protection supplies both trains of the RTS. When testing Channel 1, Train A and Train B must be examined.
Similarly, Train A and Train B must be examined when testing Channel 2, Channel 3, and Channel 4 (if applicable). The CHANNEL CALIBRATION and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required channel accuracies (Ref. 8).
SR 3.3.1.1 A CHANNEL CHECK is required for the following RTS trip functions:
- Power Range Neutron Flux-High;
- Power Range Neutron Flux-Low;
- Intermediate Range Neutron Flux;
- Source Range Neutron Flux;
- Overtemperature AT;
- Overpower AT;
- Pressurizer Pressure-Low;
- Pressurizer Pressure-High;
- Pressurizer Water Level-High;
- Reactor Coolant Flow-Low (Single Loop);
R.E. Ginna Nuclear Power Plant B 3.3.1-39 Revision 61
RTS Instrumentation B 3.3.1
" Reactor Coolant Flow-Low (Two Loops); and
" SG Water Level-Low Low Performance of the CHANNEL CHECK encooevcy 12 heaps ensures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or of more serious instrument conditions. A CHANNEL CHECK will detect gross channel failure; thus, it is a verification that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Channel check acceptance criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.
The F*rqoune.y of 12 houra i* -- based-. an .tin--.xperien* e that ehannel failur i3 raro. The CHANNEL CHEC
,iemntratcc supplements less.m1 f., , but marc irl qucnt, checks i f
.hannelo durin1g normal8 oporaitienal use of the displays asseeiated with the LCOG roqUirod ehannels.
SR 3.3.1.2 This SR compares the calorimetric heat balance calculation to the NIS Power Range Neutron Flux-High channel output e,,epy.24.heta.. If the calorimetric exceeds the NIS channel output by > 2% RTP, the NIS is still OPERABLE but must be adjusted. If the NIS channel output cannot be properly adjusted, the channel is then declared inoperable.
This SR is modified by a Note which states that this Surveillance is required to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after power is Ž_50% RTP. At lower power levels, calorimetric data are inaccurate.
The Fr-lqu y of eve; 24 huFrA i, based en plent o*pratin,*1.
eensidering@ in19tru~mnt roliability and epcraiting his9tory data forintumn drift. Tegether these factors dcmenStraltc the ehange in the abselutc difefcroncc between NIS and heat belenec ealeulated pewcrc rarcly-emeeeds 2% en any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> iporiod.
R.E. Ginna Nuclear Power Plant B 3.3.1-40 Revision 61
RTS Instrumentation B 3.3.1 In addition, control room operators periodically monitor redundant indications and alarms to detect deviations in channel outputs.
SR 3.3.1.3 JINSERT 3 This SR compares the incore system to the NIS channel output eery 31
^ffe.tiv. full p.w.. days (EFP., ). If the absolute difference is > 3%, the NIS channel is still OPERABLE, but must be readjusted. If the NIS channel cannot be properly readjusted, the channel is then declared inoperable. This surveillance is performed to verify the f(AI) input to the Overtemperature AT Function.
This SR is modified by two Notes. Note 1 clarifies that the Surveillance is required to be performed within 7 days after THERMAL POWER is> 50%
RTP but prior to exceeding 90% RTP following each refueling and if it has not been performed within the last 31 EFPD. Note 2 states that performance of SR 3.3.1.6 satisfies this SR since it is a more comprehensive test.
Thc Fr.quen.y of ..v.; 31 EFPD is based en plant . p..ating.
cxpriscc eensidering finstrumcnet rcliability and eperating histery data fcntru mcnt drift. Also, the slew ehano ctrn flux during the fuel in eyele ean be deteeted during thiS imterval.
SR 3.3.1.4 E f l This SR is the performance of a TADOT evcr; 31 days on a STACCERED TEST BASIS of the RTB, and the RTB Undervoltage and Shunt Trip Mechanisms. This test shall verify OPERABILITY by actuation of the end devices.
The test shall include separate verification of the undervoltage and shunt trip mechanisms except for the bypass breakers which do not require separate verification since no capability is provided for performing such a test at power. The independent test for bypass breakers is included in SR 3.3. 1.11. However, the bypass breaker test shall include a local shunt trip. This test must be performed on the bypass breaker prior to placing it in service to take the place of a RTB.
based en industry epcraiting cxperiencc, ccnsidcring inStrumcnt roliability and eperating histery data-.
R.E. Ginna Nuclear Power Plant B 3.3.1-41 Revision 61
RTS Instrumentation B 3.3.1 SR 3.3.1.5 This SR is the performance of an ACTUATION LOGIC TEST on the RTS Automatic Trip Logic ..... ; 31 days en a STAGCERED TEST BASIS.
The train being tested is placed in the bypass condition, thus preventing inadvertent actuation. All possible logic combinations, with and without applicable permissives, are tested for each protection function. T-he Froqueney ef eoveo; 31 days en a STACCERED TEST BASIS 09 based eon induatry epefrating expe ionc cnidering inq~trumcnt roliability and epefrating hok*tFRyF data*-.
SR 3.3.1.6 This SR is a calibration of the excore channels to the incore channels evwety 92 E -.. If the measurements do not agree, the excore channels are still OPERABLE but must be calibrated to agree with the incore detector measurements. If the excore channels cannot be adjusted, the channels are then declared inoperable. This surveillance is performed to verify the f(AI) input to the Overtemperature AT Function.
A minimum of 2 thimbles per quadrant and sufficient movable incore detectors shall be operable during recalibration of the excore axial off-set detection system. To calibrate the excore detector channels, it is only necessary that the movable incore system be used to determine the gross power distribution in the core as indicated by the power balance between the top and bottom halves of the core.
This SR has been modified by a Note stating that this Surveillance is required to be performed within 7 days after THERMAL POWER isŽ 50%
RTP but prior to exceeding 90% RTP following each refueling.
The Frogueney of 92 EFPD isadequate based ein industry eperating experuencc, considcrinig ino8trumonet roliability and epefrating histery data for inStFrumcnt drit.
SR 3.3.1.7 E f l This SR is the performance of a COT e%'e~y-92-days-for the following RTS functions:
" Power Range Neutron Flux-High;
" Source Range Neutron Flux (in MODE 3, 4, or 5 with CRD System capable of rod withdrawal or all rods not fully inserted);
" Overtemperature AT;
- Overpower AT;
- Pressurizer Pressure-Low; R.E. Ginna Nuclear Power Plant B 3.3.1-42 Revision 61
RTS Instrumentation B 3.3.1
- Pressurizer Pressurizer-High;
- Pressurizer Water Level-High;
" Reactor Coolant Flow-Low (Single Loop);
- Reactor Coolant Flow-Low (Two Loops); and
- SG Water Level-Low Low A COT is performed on each required channel to ensure the channel will perform the intended Function. The as-found setpoints must be within the COT Acceptance Criteria specified within plant procesures. The as-left values must be consistent with the setting tolerance used in the setpoint methodology (Ref. 8).
This SR is modified by a Note that provides a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> delay in the requirement to perform this surveillance for source range instrumentation when entering MODE 3 from MODE 2. This Note allows a normal shutdown to proceed without a delay for testing in MODE 2 and for a short time in MODE 3 until the RTBs are open and SR 3.3.1.7 is no longer required to be performed. If the plant is in MODE 3 with the RTBs closed for greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, this SR must be performed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ilkl*-'*r--r"'*"r-
- I after entry into MODE 3.
111mor- I a rl hp Frequeney ef 92 days is ee.nsustent with Refccnee .
SR 3.3.1.8 This SR is the performance of a COT as described in SR 3.3.1.7 for the Power Range Neutron Flux-Low, Intermediate Range Neutron Flux, and Source Range Neutron Flux (MODE 2), except that this test also includes verification that the P-6 and P-10 interlocks are in their required state for the existing plant condition. This SR is modified by two Notes that provide a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> delay in the requirement to perform this surveillance.
These Notes allow a normal shutdown to be completed and the plant removed from the MODE of Applicability for this surveillance without a delay to perform the testing required by this surveillance. The Frequency INSa p- p 92 li dayse.y applies if the plant remains in the MODE of Applicability ISR 1 jI "after the initial performances of prior to reactor startup and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reducing power below P-10 or P-6.
R.E. Ginna Nuclear Power Plant B 3.3.1-43 Revision 61
RTS Instrumentation B 3.3.1 The MODE of Applicability for this surveillance is < 6% RTP for the power range low and intermediate range channels and < 5E-1lamps for the Source range channels. Once the plant is in MODE 3, ihis surveillance is no longer required. If power is to be maintained < 6% RTP or < 5E-1lamps for more than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, then the testing required by this surveillance must be performed prior to the expiration of the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> limit, unless perorm prr2d
."-heth . Four hours is a reasonable time to complete the required testing or place the plant in a MODE where this surveillance is no longer required. This test ensures that the NIS source, intermediate, and power range low channels are OPERABLE prior to taking the reactor critical or after reducing power into the applicable MODE (< 6% RTP or < 5E-1lamps) for periods > 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SR 3.3.1.9 JINSERT 3 This SR is the performance of a TADOT for the Undervoltage-Bus 11 A and 11B and Underfrequency-Bus 11A and 11B trip Functions. T:he-Frequeney 1fevev, 92 d*ys izs -ensistnt with Rfer1ncc 9.
This SR is modified by a Note that excludes verification of setpoints from the TADOT. Since this SR applies to Bus 11A and 11B undervoltage and underfrequency relays, setpoint verification requires elaborate bench calibration and is accomplished during the CHANNEL CALIBRATION required by SR 3.3.1.10.Ax SR 3.3.1.10"L-'IJSERT3 This SR is the performance of a CHANNEL CALIBRATION for the following RTS Functions:
- Power Range Neutron Flux-High;
- Power Range Neutron Flux-Low;
" Intermediate Range Neutron Flux;
" Source Range Neutron Flux;
" Overtemperature AT;
- Overpower AT;
- Pressurizer Pressure-Low;
" Pressurizer Pressure-High;
" Pressurizer Water Level-High;
" Reactor Coolant Flow-Low (Single Loop);
R.E. Ginna Nuclear Power Plant B 3.3.1-44 Revision 61
RTS Instrumentation B 3.3.1
- Reactor Coolant Flow-Low (Two Loops);
- Undervoltage-Bus 11A and 11B;
- Underfrequency-Bus 11A and 11B;
- SG Water Level-Low Low;
- Turbine Trip-Low Autostop Oil Pressure; and
- Reactor Trip System Interlocks.
A CHANNE=L CALIBRATIGN is pcrfefrmzd evcr; 24 mcenths, er app...imat.ly at ev"*r r"fu:ling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.
CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the plant specific setpoint methodology (Ref. 8). The difference between the current as-found values and the previous test as-left values must be consistent with the drift allowance used in the setpoint methodology.
The Frcqucncy ef 24 months is based en the assumptien of 24 month ealibratifin intewersacin the detefrminatien ef the magnitude of equipment drift in the sctpeint mcethedelegy.
With respect to RTDs, whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the resistance temperature detectors (RTD) sensors shall include an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. This is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.
This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. The CHANNEL CALIBRATION for the power range neutron detectors consists of a normalization of the detectors based on a power calorimetric and flux map performed above 50% RTP. The CHANNEL CALIBRATION for the source range and intermediate range neutron detectors consists of obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. This Surveillance is not required for the NIS power range detectors for entry into MODE 2 or 1, and is not required for the NIS intermediate range detectors for entry into MODE 2, because the plant must be in at least MODE 2 to perform the test for the intermediate range detectors and MODE 1 for the power range detectors. The 24 mo.nth F- qu.n. y i. based en the need teC perfel this Su1,ViI, urlllar-.
und1r the Irnditic,, that apply during a pllnt R.E. Ginna Nuclear Power Plant B 3.3.1-45 Revision 61
RTS Instrumentation B 3.3.1
,utago and the pet.ntial .f-r 19 an*
una`ind transi-nt if the Su,.II*an
.. ,, p-"form--d with the "-*--tr Bt p"wr. Op-rating ,epe;Ciono, has sh.wn these . ^mpno*nt^ usually pass theoSure..l",n. when pefnrmod en the 24 month Frogucncy.
SR 3.3.1.11 t-tIN.SERT3 This SR is the performance of a TADOT of the Manual Reactor Trip, RCP Breaker Position, and the Sl Input from ESFAS trip Functions. This-TA.DOT is p..f....m..d evry 24 months. This test independently verifies the OPERABILITY of the undervoltage and shunt trip mechanisms for the Manual Reactor Trip Function for the Reactor Trip Breakers and Reactor Trip Bypass Breakers.
The Frcqueney is based ong the known roliability of the Functions andl the mult..hanncl r"dundanoy available, and has been shown to bo aoscptablc through oporating oxporionos.
SR 3.3.1.12 .- ER3 This SR is the performance of a TADOT for Turbine Trip Functions which is performed prior to reactor startup if it has not been performed within the last 31 days. This test shall verify OPERABILITY by actuation of the end devices.
The Frequency is based on the known reliability of the Functions and the multichannel redundancy available, and has been shown to be acceptable through operating experience.
This SR is modified by a Note stating that verification of the Trip Setpoint does not have to be performed for this Surveillance. Performance of this test will ensure that the turbine trip Function is OPERABLE prior to taking the reactor critical because portions of this test cannot be performed with the reactor at power.
SR 3.3.1.13 This SR is the pcreffcranooe of a COTF of the RTS interlooks evcr; 24 The fl I l..l*
Froguoncy l ,7Ii*
- is based ITI3i on the known roliability ofl the
- .1U
- u*.# T.
- l* l intcrlocks l V i and
[
tho mnultihanncl rcdandenoy available, and has boon shown to b acccptablc through opefrating exporionoc.
R.E. Ginna Nuclear Power Plant B 3.3.1-46 Revision 61
RTS Instrumentation B 3.
3.1 REFERENCES
- 2. 10 CFR 50.67.
- 3. American National Standard, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," N18.2-1973.
- 4. UFSAR, Chapter 7.
- 5. UFSAR, Chapter 6.
- 6. UFSAR, Chapter 15.
- 7. IEEE-279-1971.
- 8. EP-3-S-0505, "Instrument Setpoint/Loop Accuracy Calculation Methodology".
- 9. WCAP 10271 P A, Suppl^c., nt 2, Rey. 1, Jun 1990*.
I~e R.E. Ginna Nuclear Power Plant B 3.3.1-47 Revision 61
ESFAS Instrumentation B 3.3.2 SURVEILLANCE The SRs for each ESFAS Function are identified by the SRs column of REQUIREMENTS Table 3.3.2-1. Each channel of process protection supplies both trains of the ESFAS. When testing Channel 1, Train A and Train B must be examined. Similarly, Train A and Train B must be examined when testing Channel 2, Channel 3, and Channel 4 (if applicable). The CHANNEL CALIBRATION and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required charnel accuracies.
A Note has been added to the SR Table to clarify that Table 3.3.2-1 determines which SRs apply to which ESFAS Functions.
SR 3.3.2.1 This SR is the performance of a CHANNEL CHECK for the following ESFAS Functions:
SI-Containment Pressure-High; SI-Pressurizer Pressure-Low;
- SI-Steam Line Pressure-Low;
" CS-Containment Pressure-High High;
- Steam Line Isolation-Containment Pressure-High High;
" Steam Line Isolation-High Steam Flow Coincident with SI and Tavg-Low;
- Steam Line Isolation-High-High Steam Flow Coincident with SI;
- Feedwater Isolation-SG Water Level-High; and
- AFW-SG Water Level-Low Low.
Performance of the CHANNEL CHECK oncccvcry 12 heur3 ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or of more serious instrument conditions. A CHANNEL CHECK will detect gross channel failure; thus, it is a verification the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
R.E. Ginna Nuclear Power Plant B 3.3.2-31 Revision 42
ESFAS Instrumentation B 3.3.2 CHANNEL CHECK acceptance criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.
-- ! .... L nc -rcu i:Ucncv i J or3 zri;N;Z- iac CIM 1 ii~ ci-7; _-tij cxcic ZAn dcmnzrnstmte ehannelftaiurc *9 raro. I he CHANNEL CHECK supplements less fefrmal, but mcre froguent, ehccks ef ehanncls dluring nrmal8 epcraltienal use ef the displays asseenated with the LCO)rcguircd SR 33.2.2 This SR is the performance of a COT eve.y92-days for the following ESFAS functions:
- SI-Containment Pressure-High;
- SI-Pressurizer Pressure-Low;
- SI-Steam Line Pressure-Low;
- CS-Containment Pressure-High High;
- Steam Line Isolation-Containment Pressure-High High; Steam Line Isolation-High Steam Flow Coincident with SI and Tavg-Low;
- Steam Line Isolation-High-High Steam Flow Coincident with SI;
- Feedwater Isolation-SG Water Level-High; and
- AFW-SG Water Level-Low Low.
A COT is performed on each required channel to ensure the channel will perform the intended Function. Setpoints must be found to be within the COT Acceptance Criteria specified in plant procedures. The as-left values must be consistent with the drift allowance used in the setpoint methodology.
The Froquency 3f 92 days *9-eeif~sistent with ninRcfercnec 7.Th Frcguency is adequate base d-ef aindustry epefrating expcriencc, eens Edcr ME in~tFUment rcl ab litv and
-*-rat1ng hiStr dta..
[INSERT3 R.E. Ginna Nuclear Power Plant B 3.3.2-32 Revision 42
ESFAS Instrumentation B 3.3.2 SR 3.3.2.3 This SR is the performance of a TADOT evey 92 d-ays. This test is a check of the AFW-Undervoltage-Bus 11A and 11B Function.
The test includes trip devices that provide actuation signals directly to the protection system. The SR is modified by a Note that excludes verification of setpoints for relays. Relay setpoints require elaborate bench calibration and are verified during CHANNEL CALIBRATION. The-Fr-qucn.y ef 92 days is adequate based en i^ndustr; epcratin cx lcrIc1 I'*c, Icnidcring inIstIuI I nclt cliability and ltirgepl* ; datc.
SR 3.3.2.4 [INSERT 31----
This SR is the performance of a TADOT every 24 ,,e,the. This test is a check of the SI, CS, Containment Isolation, Steam Line Isolation, and AFW Manual Initiations, and the AFW-Trip of Both MFW Pumps Functions. Each Function is tested up to, and including, the master transfer re JINSERT31 perating experienee and *3eenicsitent with the typieal rcfueling eyele.
The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Manual Initiations, and AFW-Trip of Both MFW Pumps Functions have no associated setpoints.
SR 3.3.2.5 This SR is the performance of a CHANNEL CALIBRATION-eve*y 24 meths of the following ESFAS Functions:
SI-Containment Pressure-High; SI-Pressurizer Pressure-Low;
- SI-Steam Line Pressure-Low;
- CS-Containment Pressure-High High;
- Steam Line Isolation-Containment Pressure-High High; Steam Line Isolation-High Steam Flow Coincident with SI and Tavg-Low;
- Steam Line Isolation-High-High Steam Flow Coincident with SI; Feedwater Isolation-SG Water Level-High; AFW-SG Water Level-Low Low; and AFW-Undervoltage-Bus 11A and 11B.
R.E. Ginna Nuclear Power Plant B 3.3.2-33 Revision 42
ESFAS Instrumentation B 3.3.2 CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.
CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the plant specific setpoint methodology. The "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The Frcgqucney ef 24 menths as based en the assumptien ef a 24 menth ealibraktienitra n the detefrminaticn ef the maegnitude ef equipment drift in the sctpeint mnethedelegy.
SR 3.3.2.6 This SR ensures the SI-Pressurizer Pressure-Low and SI-Steam Line Pressure-Low Functions are not bypassed when pressurizer pressure
> 2000 psig while in MODES 1, 2, and 3. Periodic testing of the pressurizer pressure channels is required to verify the setpoint to be less than or equal to the limit.
The difference between the current as-found values and the previous test as-left values must be consistent with the drift allowance used in the setpoint methodology (Ref. 6). The setpoint shall be left set consistent with the assumptions of the current plant specific setpoint methodology.
If the pressurizer pressure interlock setpoint is nonconservative, then the Pressurizer Pressure-Low and Steam Line Pressure-Low Functions are considered inoperable. Alternatively, the pressurizer pressure interlock can be placed in the conservative condition (nonbypassed). If placed in the nonbypassed condition, the SR is met and the Pressurizer Pressure-Low and Steam Line Pressure-Low Functions would not be considered inoperable.
SR 3.3.2.7 I This SR is the performance of an ACTUATION LOGIC TEST on all ESFAS Automatic Actuation Logic and Actuation Relays Functions eVeiy-24 menths. This test includes the application of various simulated or actual input combinations in conjunction with each possible interlock state and verification of the required logic output. Relay and contact operation is verified by a continuance check or actuation of the end device.
The Frequeney ef 24 Fflnths is based en epcrating cxperienee and the need te perffermf thus testing duFrig a plant shutdewn te prevent 8 reaetr trip frem eeeurring.
R.E. Ginna Nuclear Power Plant B 3.3.2-34 Revision 42
PAM Instrumentation B 3.3.3 G..1 If one channel for Function 7 or 10 cannot be restored to OPERABLE status within the required Completion Time of Condition D, the plant must take immediate action to prepare and submit a special report to the NRC.
This report shall be submitted within the following 14 days from the time the action is required. This report discusses the alternate means of monitoring Reactor Vessel Water Level and Containment Area Radiation, the degree to which the alternate means are equivalent to the installed PAM channels, the areas in which they are not equivalent, and a schedule for restoring the normal PAM channels.
These alternate means must have been developed and tested and may be temporarily installed if the normal PAM channel(s) cannot be restored to OPERABLE status within the allotted time.
SURVEILLANCE A Note has been added to the SR Table to clarify that SR 3.3.3.1 and SR REQUIREMENTS 3.3.3.2 apply to each PAM instrumentation Function in Table 3.3.3-1.
SR 3.3.3.1 Performance of the CHANNEL CHECK .n.. c;c ry 31 days ensures that a gross instrumentation failure has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of more serious instrument conditions. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be compared to similar plant instruments located throughout the plant.
Channel check acceptance criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.
As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized.
The Frogueney ef 31 days is based cxpcriencc that
.. .p..ating defmeonStrates that ehannel failuro is rarc. The CHANNELI=GCHECK supplements less fefrmal, but ffiro froguent, eheeks of ehannels durin9g R.E. Ginna Nuclear Power Plant B 3.3.3-16 Revision 73
PAM Instrumentation B 3.3.3 ncral cpefzrtie S use of the display RSScmatfea with the LCO) rc1uiFd*
,,a SR3.3.3.21 A (LA NIMIrI ¢A6"IRATII N is p*A'1"1r1d ,nths,24 mec er ap...ximatcly at cv,. r'fueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to the measured parameter with the necessary range and accuracy. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the Core Exit thermocouple sensors shall include an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. This is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The F-r..u.ncy is based en .p.rating-a -ndCX as CTC*CC ensistent
.. . *,,,.,: with the typical
- ,-^ - .. ÷ ÷..^ .,*: industr" i *A.. .. rceling
- .^: . eyele.
...3 REFERENCES 1. UFSAR, Section 7.5.2.
- 2. Regulatory Guide 1.97, Rev. 3.
- 3. NUREG-0737, Supplement 1, "TMI Action Items."
- 4. UFSAR, Section 6.2.5.
R.E. Ginna Nuclear Power Plant B 3.3.3-17 Revision 73
LOP DG Start Instrumentation B 3.3.4 significantly reduce the probability that the LOP DG start instrumentation will trip when necessary.
SR 3.3.4.1 This SR is the performance of a TADOT eyeoy 3-1days. This test checks trip devices that provide actuation signals directly. For these tests, the relay trip setpoints are verified and adjusted as necessary to ensure the LSSS can still be met. Thc 31 day FFr..u.n.y i. based en the kn.wn roliability of the r-elays and eentrols and has been shewn to be aeecptable thr.ugh ope,, ting
,xporioene'. IN E T -
SR 3.3.4.2 3 This SR is the performance of a CHANNEL CALIBRATION evesy 24 months, r appr...fimat"ly at .v..y fu'ling, of the LOP DG start instrumentation for each 480 V bus.
The voltage setpoint verification, as well as the time response to a loss of voltage and a degraded voltage test, shall include a single point verification that the trip occurs within the required time delay.
CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.
The of 24 .. nths i. based on epcrating cxporicncc
.F..u.ncy eensmstent with the typieal industry rofucling eyelc and is justified by the assufmption of a 24 menth ealibffltift intewe'l win the detrmqinaltien of the mnagnitude of equipment drift in the setpeint analysi.
REFERENCES 1. UFSAR, Section 8.3.
I 2. UFSAR, Chapter 15.
R.E. Ginna Nuclear Power Plant B 3.3.4-7 Revision 37
Containment Ventilation Isolation Instrumentation B 3.3.5 SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.5-1 REQUIREMENTS determines which SRs apply to which Containment Ventilation Isolation Functions.
SR 3.3.5.1 Performance of the CHANNEL CHECK cnccc;vcr; 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ensures that a gross failure of instrumentation has not occurred and the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The CHANNEL CHECK agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. Ifa channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.
The Frogueney isbased en eperating- cxcI !that demenstmtcs but moroe frgucnt, cheeks of ehannels during nrmFFal operational use of the displays asseeiated with the LCOG Fcqu ird channcls.
SR 3.3.5.2 A COT is performed e.e.y .92 ays-.on each required channel to ensure the channel will perform the intended Function. The Frequency is based on the staff recommendation for increasing the availability of radiation monitors according to NUREG-1 366 (Ref. 2). This test verifies the capability of the instrumentation to provide the containment ventilation system isolation. The setpoint shall be left consistent with the current plant specific calibration procedure tolerance.
SR 3.3.5.3 k-.INS E.R1T 3I This SR is the performance of an ACTUATION LOGIC TEST. All possible logic combinations, with and without applicable permissives, are tested for each protection function. In addition, the master relay is tested for continuity. This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relay coils. Th'is test i perf.rmcd e,-,v; 24 months. The Su.v,*ilanc1 interval -s aee.ptabl-based on. inStrun*,,
,, t liability and indust,;y operiting -xp-*- ne-"-.
4\
JINSERT 3 R.E. Ginna Nuclear Power Plant B 3.3.5-8 Revision 42
Containment Ventilation Isolation Instrumentation B 3.3.5 SR 3.3.5.4 A ....ANN.I=. GAI...ATII as... pef...mt..d cvcr' 24 month** e appr..imatcly at .. v... . .fueling.
CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.
The Fr,**,nIy as based enIp rtnFgt cxIcII nc and I9 cns.st.nt with the typieel industry Fefueling eyele.
REFERENCES 1. 10 CFR 50.67.
- 2. NUREG-1 366.
R.E. Ginna Nuclear Power Plant B 3.3.5-9 Revision 42
CREATS Actuation Instrumentation B 3.3.6 C.1 and C.2 Condition C applies when the Required Action and associated Completion Time of Condition A or B has not been met and the plant is in MODE 1, 2, 3, or 4. The plant must be brought to a MODE that minimizes accident risk. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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.
D.1 and D.2 Condition D applies when the Required Action and associated Completion Time of Condition A or B has not been met during movement of irradiated fuel assemblies. Movement of irradiated fuel assemblies must be suspended immediately to reduce the risk of accidents that would require CREATS actuation. This places the plant in a condition that minimizes risk. This does not preclude movement of fuel or other components to a safe position.
SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.6-1 REQUIREMENTS determines which SRs apply to which CREATS Actuation Functions.
SR 3.3.6.1 Performance of the CHANNEL CHECK cncccvcry 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> cnsures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or of more serious instrument conditions. A CHANNEL CHECK will detect gross channel failure; thus, it is a verification that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
CHANNEL CHECK acceptance criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.
The Frcgucr-ey ef 12 heur3 is based en epcrotirng cxpcriencc h dcmcneS4trotc ehennel failuro iSroro. The CHANNEL CHEC supplements less feFrmal, but moroe froguont, chocks of eharnigeS during R.E. Ginna Nuclear Power Plant B 3.3.6-7 Revision 38
CREATS Actuation Instrumentation B 3.3.6 nor al op ,rationaluse
, f the displays ess,,iatd with the I r'" , ,quir*
SR33.6.2~T This SR is the performance of a COT onco c'er;y 92 days on each required channel to ensure the channel will perform the intended function. This test verifies the capability of the instrumentation to provide the automatic CREATS actuation. The setpoints shall be left consistent with the plant specific calibration procedure tolerance. Thc Frqu.ncy
- f 92 days is based en the knoiwn roliabiliety of the monitoring equipment and has been shewn to be acooptablo througoh epeffiting epeoneeo.\
SR 3.3.6.3 NSERT 3 This SR is the performance of a TADOT of the Manual Initiation Function eve.* 24.mentl.s. The Manual Initiation Function is tested up to, and including, the master relay coils.
The Froguonoy of 24 months is based on the 1(noWn Foliability of the Funetion and the rodundancy available, and has boon shown to be aeooptablo... ,,. ......
through ..,,,ok .. ..
,. .. ... experienec.
oporating iNSERT 31 The SR is modified by a Note that excludes verification of setpoints because the Manual Initiation Function has no setpoints.
SR 3.3.6.4 This SR is the performfanoc ef a CHANNEL GALlBRATION c-vcry 24 m.nths, or appr,..imat.ly at c..;y . .fu.ling.
CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.
The Frogquency of 24 months is based on epcraltinq expcrienee and i ocrnsistont with the typical industn Fcfucling eyeloe SR 3.3.6.5 L'INSERT This SR is the performance of an ACTUATION LOGIC TEST. All possible logic combinations are tested for the CREATS actuation instrumentation. In addition, the master relay is tested forcontinuity. This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relay coils. This test is acceptable based on instrument reliability and operating experience.*
t3 R.E. Ginna Nuclear Power Plant B 3.3.6-8 Revision 38
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation. If the indicated RCS total flow rate is below the LCO limit, power must be reduced, as required by Required Action B.1, to restore DNB margin and eliminate the potential for violation of the accident analysis bounds.
The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time for restoration of the parameters provides sufficient time to determine the cause for the off normal condition, to adjust plant parameters, and to restore the readings within limits, and is based on plant operating experience.
B._1 If Required Action A.1 is not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply.
To achieve this status, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis bounds. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable to reach the required plant conditions in an orderly manner.
SURVEILLANCE SR 3.4.1.1 REQUIREMENTS Sinec Requircd Action A.!1 allcws a Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to rostero parametcrS that arc net within limits, the 12 hcur Guryeillanee Frcequcncy fer PrcssUri~zer prcssurce Is sufficiont to cncSUrc the prcSSUrc can be rosterod to a normafll cpcratien, steady state eendition fellewing lead eh".g..s and ether .xpc.ted transient p.r*ations. Thc 12 heur .ntr..al-has been shown by opcralting pracetics to be sufficicnt to rcgula rly assess for pctential degradatien and tc Yerify opefrtieon is within safety analysis SR 3.4.1.2 NSERT3 Sinec Rcquircd Aetion A.! allows a Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> tc Festerc pa....mteS that are not within knmits, thc 12 hu" .Su.vCiI an.. Frc.u.n.y for RCS ..... g. tc..p..atur' is suff'*1*
- nt to nsurc the t"mpleat*c* can be.. tf-cd to a norm.al p...ati-n, steady statc ccnditien following load changcs and othcr expcctcd tranfsicnit epcratiens. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> intcrwal has been shown by opcrating practicc to be sufficicnt to rcgulafly assess a.#s, I m pt,*I e n1 s,.r't A II.#
.. 11 J . l .ll.lI for potential degradation and to reif, epefration is within safety anaelysis
.#.11 11 1*1I I IIV l I Ir. ,l R.E. Ginna Nuclear Power Plant B 3.4.1-4 Revision 42
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 SR 3.4.1.3 Measurement of RCS total flow rate oncc cvcry 24 months verifies the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate. This verification may be performed via a precision calorimetric IINSERT 3 F----.7"--. ... heat
.. balance
-. ^,. or other accepted-. means.
^1,.*,. ,.,. . .*.. *.^ ,..
after a rofucling eutagc when the ccrc has been alterod, whieh maey have
.aus.d an altcratine ef f..W rcc.tancc.. Verification of RCS flow rate on a shorter interval is not required since this parameter is not expected to vary during steady state operation as there are no RCS loop isolation valves or other installed devices which could significantly alter flow.
Reduced performance of a reactor coolant pump (RCP) would be observable due to bus voltage and frequency changes, and installed alarms that would result in operator investigation.
This SR is modified by a Note that allows entry into MODE 1, without having performed the SR, and placement of the plant in the best condition for performing the SR. The Note states that the SR shall be performed within 7 days after reaching 95% RTP. This exception is appropriate since the heat balance requires the plant to be at a minimum of 95% RTP to obtain the stated RCS flow accuracies.
REFERENCES 1. UFSAR, Chapter 15.
- 2. NRC Memorandum from E.L. Jordan, Assistant Director for Technical Programs, Division of Reactor Operations Inspection to Distribution;
Subject:
"Discussion of Licensed Power Level (AITS F14580H2)," dated August 22, 1980.
R.E. Ginna Nuclear Power Plant B 3.4.1-5 Revision 42
RCS Minimum Temperature for Criticality B 3.4.2 ACTIONS A.1 If the parameters that are outside the limit cannot be restored, 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 MODE 2 with Keff < 1.0 within 30 minutes. Rapid reactor shutdown can be readily and practically achieved within a 30 minute period due to the proximity to MODE 2 conditions.
The allowed time is reasonable, based on operating experience, to reach MODE 2 with Keff < 1.0 in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.4.2.1 REQUIREMENTS This SR verifies that RCS Tavg in each loop is > 540OF within 30 minutes prior to achieving criticality. This ensures that the minimum temperature for criticality is being maintained just before criticality is reached. The 30 minute time period is long enough to allow the operator to adjust temperatures or delay criticality so the LCO will not be violated, thereby providing assurance that the safety analyses are not violated.
SR 3.4.2.2 RCS loop average temperature is required to be verified at or above 540OF every 30 minutes in MODE 1, and in MODE 2 with keff __1.0. The 30 minute frequency is sufficient based on the low likelihood of large temperature swings without the operators knowledge.t-lNSERT 3i This SR is modified by a Note that only requires the SR to be performed if any RCS loop Tavg is < 547 0 F and the low Tavg alarm is either inoperable or not reset. The Tavg alarm provides operator indication of low RCS temperature without requiring independent verification while a Tavg
> 547 0 F in both RCS loops is within the accident analysis assumptions. If the Tavg alarm is to be used for this SR, it should be calibrated consistent with industry standards.
This surveillance is replaced by SR 3.1.8.2 during PHYSICS TESTING.
REFERENCES 1. None.
R.E. Ginna Nuclear Power Plant B 3.4.2-3 Revision 21
RCS P/T Limits B 3.4.3 SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the PTLR limits is required eveFy--30 nlntkea-when RCS pressure and temperature conditions are undergoing planned changes. This Freque- i.er asne in view the contrOl room findication avaF-ila~ble fto monintor ROS status. AlsIo, sfin termperaturo rate of change limfitS are specified in haurly inecroments, 30 minutits permlitS assessment and corroctien for mninor deviations within a
. .easenable Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.
This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.
REFERENCES 1. WCAP-14040, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 1, December 1994.
- 3. ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.
- 4. ASTM E 185-82, July 1982.
- 6. Regulatory Guide 1.99, Revision 2, May 1988.
- 7. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.
R.E. Ginna Nuclear Power Plant B 3.4.3-6 Revision 21
RCS Loops - MODE 1 > 8.5% RTP B 3.4.4 Operation in other MODES is covered by:
LCO 3.4.5, "RCS Loops - MODES 1
- 8.5% RTP, 2, AND 3";
LCO 3.4.6, "RCS Loops - MODE 4";
LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled";
LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled";
LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-Water Level > 23 Ft" (MODE 6); and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Water Level < 23 Ft" (MODE 6).
ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the plant to MODE 1 < 8.5% RTP. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits.
The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 1 < 8.5% RTP from full power conditions in an orderly manner and without challenging safety systems.
SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verification eveiy 12 het'. that each RCS loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal while maintaining the margin to DNB. Use of control board indication for these parameters is an acceptable verification. ThecF..u.n.y.f 12 heura is sufficicnt eensidering ether indicotiecns and 818Frm3 available te t,-he epe tc in thec.... ,nr r.... tm m - - -^itorCS loop p.f... la ... .*
INSERT 3 R.E. Ginna Nuclear Power Plant B 3.4.4-3 Revision 46
RCS Loops - MODES 1 _<8.5% RTP, 2, and 3 B 3.4.5 B.1 If restoration of the inoperable loop is not possible within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the plant must be brought to MODE 4. In MODE 4, the plant may be placed on the Residual Heat Removal System. The additional Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is compatible with required operations to achieve cooldown and depressurization from the existing plant conditions in an orderly manner and without challenging plant systems.
C.1. C.2, and C.3 If two RCS loops are inoperable, or no RCS loop is in operation, except during conditions permitted by the Note in the LCO section, all CRDMs must be de-energized by opening the RTBs or de-energizing the MG sets. All operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended, and action to restore one of the RCS loops to OPERABLE status and operation must be initiated. Boron dilution requires forced circulation for proper mixing, and opening the RTBs or de-energizing the MG sets removes the possibility of an inadvertent rod withdrawal. Suspending the introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.
SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verification e-'ey 12 het .that the required RCS loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.
Use of the control board indication for these parameters is an acceptable verification. Th* Fc..qu.n. y.f 12 heur, *.n.idering
- 0. suffi.i.nt .the.
0,ndi;at*.in and olIrmH available tI the epeF.af in the e.otrol room t-mcne fitcr R GS lcep pcrfefrmonco.
I R.E. Ginna Nuclear Power Plant B 3.4.5-5 Revision 61
RCS Loops - MODES 1 < 8.5% RTP, 2, and 3 B 3.4.5 SR 3.4.5.2 This SR requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is
>_ 16% for two RCS loops. If the SG secondary side narrow range water level is < 16%, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink for removal of reactor or decay heat. The 12 h".. F..qucn.y is. .nsidercd adequate in view
" f ether indiesticns availableoin the eeontrcl rcen te alert the epeffitc te a less ef GGSC -
SR 3.4.5.3 Verification that the required RCP is OPERABLE ensures that an additional RCP can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump that is not in operation. The F..qu.n.y.f .7 days ... as sidc" reasenable in view of other adminiztrnt~aiy contre's available and has been shewn te be acccptable by cpcralting cxperienee.
REFERENCES 1. UFSAR Section 15.1.5.
- 2. UFSAR Section 15.4.3.
- 3. Letter from D. M. Crutchfield, NRC, to J. E. Maier, RG&E,
Subject:
"SEP Topic XV-9, Startup of an Inactive Loop, R. E. Ginna," dated August 26, 1981.
- 4. UFSAR Sections 14.6.1.5.6 and 15.2.5.
- 5. UFSAR Section 14.6.1.5.5.
R.E. Ginna Nuclear Power Plant B 3.4.5-6 Revision 61
RCS Loops - MODE 4 B 3.4.6 SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification eveFy 12 heufs that one RCS or RHR loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.
Use of control board indication for these parameters is an acceptable verification. Th* F-r..ue...y.f 12 heur- -. is suffi...nt .thc, nsidci*.g
%,nd:^atins and alarS
.. a;ilablc tc the..... ater in the... t.el rce.. te 4-
.AA ROS and RHR leep pcr efffienee.
SR 3.4.6.2 This SR requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is
> 16%. If the SG secondary side narrow range water level is < 16%, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink necessary for removal of decay heat.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frogueney as eensidc rcd adecquate in Yiew ef ether indicati'ns availabl in the "contrcl rcom tc al'^t the.p..ratcr t thc*....l SR 3.4.6.3 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump that is not in operation. The Fr**,*,ney ef 7 days is-consodcred rcasenablc in view ef ether admin~iStrativc eentrcls available and has bec shew, t. bc e ,,.ptabl"by _pcratitng xpcrienCC. 4 JINSERT 3P REFERENCES 1. UFSAR, Section 14.6.1.2.6.
R.E. Ginna Nuclear Power Plant B 3.4.6-5 Revision 61
RCS Loops - MODE 5, Loops Filled B 3.4.7 SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification every 12 haet'r-s-that one RHR loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.
Use of control board indication for these parameters is an acceptable verification. The F...quen.y of 12 hurFS is suffi.int ensidc;ing
,thor
-indcations and a'.ams available to the eperator On the e..tr.l r-cm to moenitor RHR loop perfcrmanec.
SR 3.4.7.2 t NE T3 This SR requires verification of SG OPERABILITY. Verifying that at least one SG is OPERABLE by ensuring its secondary side narrow range water level is > 16% ensures an alternate decay heat removalmethod in the event that the second RHR loop is not OPERABLE. If both RHR loops are OPERABLE, this Surveillance is not needed. Thea 12 h4e' Frequ*n*y is eensiderod adequate -tho an view of i,,
atndiopns available in the control rc. m to Ial. the o*por8atl to the less of IC IYE4L SR 3.4.7.3 [INSERT 3 Verification that a second RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the standby RHR pump. If secondary side water level is > 16% in at least one SG, this Surveillance is not needed. The Fro,,u*,ny f* 7 days is o d
,nsidcr' roasonablo in view of othor administrativ. ceotrelS aailablo and has been shown to be a...ptablc by epcr..ting .xpcrinco.
REFERENCES 1. UFSAR, Section 14.6.1.2.6
- 2. NRC Information Notice 95-35 R.E. Ginna Nuclear Power Plant B 3.4.7-5 Revision 61
RCS Loops - MODE 5, Loops Not Filled B 3.4.8 SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification ev'eiy 12 hhe's-that one RHR loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.
The Froquoeny ef 12 houro is suffiiont i cnsidering other indldatiens and a'a, H available tek, the epf.e ' the.. ee' . ' .... te....
OR.. "t^ RHR '^^p Pe~f6efafee.,ý.
SR 3.4.8.2 /INSERT3 Verification that a second RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the standby pump. The .Fr...u.nyef 7 days is ".nsiderod roasenab.' in view
.f ethcr administativ-... tr-ls available and has be.n shown tc bc REERptNblE Nponeting by expcricnee.
REFERENCES 1. None.
R.E. Ginna Nuclear Power Plant B 3.4.8-4 Revision 61
Pressurizer B 3.4.9 B.1 and B.2 If the pressurizer heaters capacity is < 100 KW, the ability to maintain RCS pressure to support natural circulation may no longer exist. By maintaining RCS pressure control, a margin to subcooling is provided.
The value of 100 KW is based on the amount needed to support natural circulation after accounting for heat losses through the pressurizer insulation during an extended loss of offsite power event.
If the capacity of the pressurizer heaters is not within the limit, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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 3.4.9.1 REQUIREMENTS This SR requires that during steady state operation, pressurizer level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The Surveillance is performed by observing the indicated level. The Fr..uen.y of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has been shown by
,per..ting praIti1e to be sufficient tO regulafly assess 'eve' fOr any deviation and verify that operation is within safoty analysos assumptions. Alarms are also available for early detection of abnormal level indications.
SR 3.4.9.2 This SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power required. This may be done by testing the power supply output by verifying the electrical load on Buses 14 and 16 with the respective heater groups on and off. The-Frequeney-of 92 days is considerod adequate to deteet heater degradation and has been show by pfoating experience to be a,,eptable./I\ ,
R.E. Ginna Nuclear Power Plant B 3.4.9-4 Revision 21
Pressurizer PORVs B 3.4.11 SURVEILLANCE SR 3.4.11.1 JINSER- 3 REQUIREMENTS Block valve cycling verifes that the valve(s) can be closed if needed. The I basis fer the F-.quc..y .f 92 days is the ASME Cod- (Ref. 2)*. Ifthe block valve is closed to isolate a PORV that is OPERABLE and is not leaking in excess of the limits of LCO 3.4.13, "RCS Operational LEAKAGE," then opening the block valve is necessary to verify that the PORV can be used for manual control of reactor pressure. Ifthe block valve is closed to isolate an otherwise inoperable PORV, the maximum Completion Time to restore the PORV and open the block valve is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, which is well within the allowable limits (25%) to extend the block valve Frequency-ef92--dxys. Furthermore, these test requirements would be completed by the reopening of a recently closed block valve upon restoration of the PORV to OPERABLE status (i.e., completion of the Required Actions fulfills the SR).
The Note modifies this SR by stating that it is not required to be performed with the block valve closed per LCO 3.4.13. This prevents the need to open the block valve when the associated PORV is leaking > 10 gpm creating the potential for a plant transient.
SR 3.4.11.2 This SR requires a complete cycle of each PORV using the nitrogen accumulators. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR. T-he-Frcqueney of 24 molnths is based en a typical rcfucling eyelc and industry eeeoptcd PFeie REFERENCES 1. UFSAR, Section 15.2.
- 2. ASME Code for Operation and Maintenance of Nuclear Power I Plants.
R.E. Ginna Nuclear Power Plant B 3.4.11-7 Revision 58
LTOP System B 3.4.12 disabling of a charging pump is necessary since RV 203 cannot mitigate a charging/letdown mismatch event if RHR is providing decay heat removal above MODE 5 and three charging pumps are operating.
The passive vent must be sized _>1.1 square inches to ensure that the flow capacity is greater than that required for the worst case mass input transient reasonable during the applicable MODES. This action is needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel and to protect the RHR system from overpressurization.
The Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to depressurize the RCS and establish a vent considers the time required to place the plant in this Condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements.
SURVEILLANCE SR 3.4.12.1 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, all SI pumps must be verified incapable of injecting into the RCS when the PORVs provide the RCS vent path (LCO 3.4.12.a) and a minimum of two SI pumps must be verified incapable of injecting into the RCS when the RCS is depressurized and an RCS vent > 1.1 square inches is established (LCO 3.4.12.b). The SI pumps are rendered incapable of injecting into the RCS through removing the power from the pumps by racking the breakers out under administrative control. An alternate method of LTOP control may be employed using at least two independent means to prevent a pump start such that a single failure or single action will not result in an injection into the RCS. This may be accomplished through the following:
- a. placing the pump control switch in the pull-stop position and closing at least one valve in the discharge flow path;
- b. locking closed a manual isolation valve in the injection path; or
- c. closing a motor operated isolation valve in the injection path and removing the AC power source.
The flowpaths through the test connections associated with the ECCS accumulator check valves (i.e., lines containing air operated valves 839A, 839B, 840A, and 840B) and the ECCS accumulator fill lines (i.e., lines containing air operated valves 835A and 835B) do not have to be isolated for this SR since the potential mass addition from a single SI pump through these six lines is limited by the installed orifices to less than that assumed for the charging/letdown mismatch analysis.
R.E. Ginna Nuclear Power Plant B 3.4.12-10 Revision 52
LTOP System B 3.4.12 The ECCS accumulator motor operated isolation valves can be verified closed by use of control board indication for valve position. This verification is only required when the accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in the PTLR. If the accumulator pressure is less than this limit, no verification is required since the accumulator cannot pressurize the RCS to or above the PORV setpoint.
The Froqueney of 12 heurs is suffleient, eensidoring other indicationis and-alars available to the eperator in th-control ro-:om, to Ycrif; the Fequir.d status of the cguiprnent. The Froquency of eyer; 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thecroafter fer GR 3 4 12 3 enGUre that the- EGGfS aeu ijmlteF nete e~r~ier~atnd mselatirm
-valves are maintained elose d and de noet rosult in a petcntial L-TOP SetRatm°n.21ý S R 3.4. 1 INSERT See SR 3.4.12.1 SR 3.4.12.3 See SR 3.4.12.1 SR 3.4.12.4 The RCS vent of > 1.1 square inches is proven OPERABLE by verifying its open condition eitheR.*
- a. Oncc evcr; 12 hourS for a vent (e.g., valve) that cannot be lockcd.
- b. Onco cvcr; 31 days for a vent (e.g., Yalyc) that ic looked sealed, or sccurcd in positien. A r1*1em d 1ros11riz1 eafety volv fits this I The passive vent arrangement must be > 1.1 square inches and be open to be OPERABLE. This Surveillance is required to be performed if the vent is being used to satisfy the pressure relief requirements of the LCO 3.4.12.b.
SR 3.4.12.5 The PORV block valve must be verified open evoey 72 hoet sto provide the flow path for each required PORV to perform its function when actuated. The valve may be remotely verified open in the main control room. This Surveillance is performed if the PORV satisfies the LCO.
The block valve is a remotely controlled, motor operated valve. The power to the valve operator is not required to be removed, and the manual operator is not required to be locked in the inactive position.
Thus, the block valve can be closed in the event the PORV develops R.E. Ginna Nuclear Power Plant B 3.4.12-11 Revision 52
LTOP System B 3.4.12 excessive leakage or does not close (sticks open) after relieving an overpressure situation.
The 72 heur Frcqueney is censidercd adequate inview ef ethr admFinistratic* ... t. 'r available t. the epolratoer in thc eentrcl reem, such als valve pesitien indieation, that Yerify that thoe PORY bleek valve rcmnains epen. T -- INSERT 31 SR 3.4.12.6 Performance of a CHANNEL OPERATIONAL TEST (COT) is required e.e.y. dys-on *,1 each required PORV to verify and, as necessary, adjust its lift setpoint. The COT will verify the setpoint is within the allowed maximum limits in the PTLR. PORV actuation could depressurize the RCS and is therefore not required.
A Note has been added indicating that this SR is required to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after decreasing RCS cold leg temperature to less than or equal to the LTOP enable temperature specified in the PTLR if it has not been performed the pr,;ieus 31 days. Depending on the cooldown rate, the CO ay not have been performed before entry into the LTOP MODES. The est must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering the LTOP MODES. he 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> considers the unlikelihood of a low temperature overpressu event during this time.
SR 3.4.12.7 FINSERT 1 Verification once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and every 31 days-thereafter that power is removed from each ECCS accumulator motor operated isolation valve ensures that at least two independent actions must occur before the accumulator is capable of injecting into the RCS. "-iee peweP-le-romeyed under administrative control 81nd -valve position is Yerificd cvcry 12 hourS, the peorfflrmnec of this surwcillanee enco within 12 heurs and cvcr; 31 days thercafter will proevide assur-ancc that peweFrIis cmovcd.
This SIR is modified by a Note which states that the Surveillance is only required when the accumulator pressure is greater than or equal to the maximum RCS pressure for the existing cold leg temperature allowed in the PTLR. If the accumulator pressure is below this limit, the LTOP limit cannot be exceeded and the surveillance is not required.
SR 3.4.12.8 ISR Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required eveFy--24 mfenths to adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known REFERENCES 1. 10 CFR 50, Appendix G.
R.E. Ginna Nuclear Power Plant B 3.4.12-12 Revision 52
IDeleted LTOP System B 3.4.12
- 2. Gencrie Lettcr8 1 "NRC Posfiticn on Em~brittlement cf Reaeter VesselI Meaetria 81nd its Impaet en Plant I r*aII tiens."
- 3. UFSAR, Section 5.2.2.
- 4. 10 CFR 50, Section 50.46.
- 6. Letter from D. L. Ziemann, NRC, to L. D. White, RG&E,
Subject:
"Issuance of Amendment No. 28 to Provisional Operating License No. DPR-1 8," dated July 26, 1979.
- 7. Generic Letter 90-06, "Resolution of Generic Issue 70, "Power-Operated Relief Valve and Block Valve Reliability," and Generic Issue 94, "Additional Low-Temperature Overpressure Protection for Light-Water Reactors."
R.E. Ginna Nuclear Power Plant B 3.4.12-13 Revision 52
RCS Operational LEAKAGE B 3.4.13 valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.
ACTIONS A..1 Unidentified LEAKAGE or identified LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.
B.1 and B.2 If any RCS pressure boundary LEAKAGE exists, or primary to secondary LEAKAGE is not within limits, or if the Required Action of Condition A cannot be completed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.
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.
In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.
SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE which is not allowed by this LCO, would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.
The RCS water inventory balance must be met with the reactor at steady state operating conditions (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows). The Surveillance is modified by two Notes. Note 1 states that this SR is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides R.E. Ginna Nuclear Power Plant B 3.4.13-4 Revision 52
RCS Operational LEAKAGE B 3.4.13 sufficient time to collect and process all necessary data after stable plant conditions are established.
Steady state operation is required to perform a proper inventory balance; calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and volume control tank levels, makeup and letdown, and RCP seal injection and return flows.
An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation."
Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.
The 7*2 heur F-,, uee,,y is a ea*e-nabl int**v*l t* trnd ILE-AGAE and rcCegngizes the impcrtoncc ef corly leakage. de~teetken inthe ffevnti en ef aeedentR.2 SR 3.4.132 INSERT3 This SR verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.17, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.
The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.
The Sure."llln FFr ,u.ncy ,f7-2 h-e..- - r. - iabl- int"-val t- tr ,nd primfary te secondar; LEAKAGE and rcccgnizes the impectanee ef carly leakage dcteetion in the provcntien ef eeeidcnts. The primarfy tee seeendar; LEFAKACE= is detefrminod using eentinueuo prccooa radiateion R.E. Ginna Nuclear Power Plant B 3.4.13-5 Revision 52
RCS Operational LEAKAGE B 3.4.13 meniteS 6r radieehcm.ieal grabo8.nampling in a"eerdancc with thc EPRI guidelines (Refcrcgeeýn).
REFERENCES 1. Atomic Industry Forum (AIF) GDC 16, Issued for comment July 10, 1967.
- 2. Generic Letter 84-04, "Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops."
- 3. UFSAR, Chapter 15.
- 4. NEI 97-06, Steam Generator Program Guidelines
- 5. EPRI, Pfessurized Water Rea~ete. PrmleIy t= Se Loa(i
,nday R.E. Ginna Nuclear Power Plant B 3.4.13-6 Revision 52
RCS PIV Leakage B 3.4.14 Required Action A.2 specifies that the double isolation barrier of two valves be restored by closing some other valve qualified for isolation.
The use of a valve other than the previously leaking PIV must include consideration that the plant may no longer be in an analyzed condition.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time after exceeding the limit considers the time required to complete the Action and the low probability of a second valve failing during this time period.
B.1 and B.2 If leakage cannot be reduced, the system isolated, or the other Required Actions accomplished, the plant must be brought to a MODE in which 1he requirement does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This Action may reduce the leakage due to reduced RCS pressure while reducing the potential for a LOCA outside the containment. 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 3.4.14.1 REQUIREMENTS Performance of leakage testing on each RCS PIV or isolation valve used to satisfy Required Action A.1 and Required Action A.2 is required to verify that leakage is below the speciled limit and to identify each leaking valve. The leakage limit of 0.5 gpmper inch of nominal valve diameter up to 5 gpm maximum applies to each valve and should be based on an RCS pressure of +/- 20 psig of normal system operating pressure.
Leakage testing requires a stable pressure condition.
For multiple in-series PIVs, the leakage requirement applies to each valve individually, except as noted below, and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other in-series valve meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.
R.E. Ginna Nuclear Power Plant B 3.4.14-5 Revision 58
RCS PIV Leakage B 3.4.14 The SI hot leg injection lines are each configured with two check valves and a motor operated valve in series. Each of these components independently is considered a qualified pressure boundary. The two check valves function as a single pressure isolation barrier and the motor operated valve serves as the second pressure isolation barrier to prevent an intersystem LOCA. Both barriers need to be tested. Testing of the check valves (877A, 877B, 878F, and 878H) and the motor operated valves (878A and 878C) identified as PIVs in the SI hot leg injection lines is to be performed at least once every 40 months. This surveillance interval is allowed since the two SI hot leg injection lines are maintained closed to address pressurized thermal shock (PTS) concerns (Ref. 7 and Ref. 11).
Testing ^f the RCS ,IVsOn #ic GI cld lcg ::cctien. lin.. an"d RHR system N to be pf..*.emfd .... y 24,months, a typical rofu'ling
' y.l. . The 24 eentefined in the Insefviee Testing Programf, ic within. the frogueney allowed by the American Seeioty ef Meeheinicol Enginecr3 (ASMVE) Code-,
I (Ref. 9), and i. based en the need to peform . u h sur-illane-- undor the eendlieion that apply durfing an eutage and the petential fr BAn unplanned trnnsient if the Suryelllanee wero perferrncd with the roaeter at-peweF.
In addition to the periodic testing requirements, testing must be performed once after the valve has been opened by flow, exercised, or had maintenance performed on it to ensure tight reseating. This maintenance does not include minor activities such as packing adjustments which do not affect the leak tightness of the valve. PIVs disturbed in the performance of this Surveillance should also be tested unless documentation shows that an infinite testing loop cannot practically be avoided. Testing must be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the valve has been reseated. A limit of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is a reasonable and practical time limit for performing this test after opening or reseating a valve.
The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.
Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance.
SR 3.4.14.2 See SR 3.4.14.1 R.E. Ginna Nuclear Power Plant B 3.4.14-6 Revision 58
RCS PIV Leakage B 3.4.14 REFERENCES 1. 10 CFR 50.2.
- 2. 10 CFR 50.55a(c).
- 4. WASH-1400 (NUREG-75/014), "An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," Appendix V, October 1975.
- 5. NUREG-0677, "The Probability of Intersystem LOCA: Impact Due to Leak Testing and Operational Changes," May 1980.
- 6. Generic Letter, "LWR Primary Coolant System Pressure Isolation Valves," dated February 23, 1980.
- 7. Letter from D. M. Crutchfield, NRC, to J. E. Maier, RG&E,
Subject:
"Order for Modification of License Concerning Primary Coolant System Pressure Isolation Valves," and associaled SER on Primary Coolant System Pressure Isolation Valves (WASH-1400, Event V),
dated April 20, 1981. (ML010542030)
- 8. EG&G Report, EGG-NTAP-6175.
I~el-toe'e-ed
- 9. ASME Ccde feF Gpcraitien and Maintemnenc ef Nuelefr Pewcr-
- 10. 10 CFR 690.*55f).
- 11. Letter from D. M. Crutchfield, NRC, to J.E. Maier, RGE,
Subject:
"TMI-2 Category "A" Items" and associated SER for Amendment No. 42 to Provisional Operating License No. DPR-18, dated May 11, 1981. (ML010540356)
R.E. Ginna Nuclear Power Plant B 3.4.14-7 Revision 58
RCS Leakage Detection Instrumentation B 3.4.15 Completion Time ensures that the plant will not be operated in a reduced configuration for a lengthy period of time.
E.1 and E.2 If a Required Action of Condition A, C, or D cannot be met, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 wilhin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 With all required monitors inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required.
SURVEILLANCE SR 3.4.15.1 REQUIREMENTS This SR requires the performance of a CHANNEL CHECK of the containment atmosphere radioactivity monitors. The check gives reasonable confidence that the channels are operating properly. T-he-Froqueney 3f 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is based en inStrumonet roliability and is-reasenable for deteoting 3ff nRmF1al eonditions.
SR 3.4.15.2 3 This SR requires the performance of a CHANNEL OPERATIONAL TEST (COT) on the containment atmosphere radioactivity monitors. The test ensures that the monitors can perform their function in the desired manner. The test verifies the alarm setpoint and relative accuracy of the instrument string Th Tc. F quny .. of 92 days.. nsidc.. S nStruf...nt
. and
.liabil.it,, p... ting cxpori.noC has Shown. thalt it is propor fer SR 3.4.15.3 &-'INSERT 31 These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Fr.quonc.y.f 24 mnths considers ehegnnl rcliabifilty and has pfvIn that IpIrting lxpo*riee this Fr.qu.n.y is . a: pt.... {e R.E. Ginna Nuclear Power Plant B 3.4.15-5 Revision 62
RCS Leakage Detection Instrumentation B 3.4.15 SR 3.4.15.4 See SR 3.4.
15.3 REFERENCES
- 3. IE Bulletin No. 80-24, "Prevention of Damage Due to Water Leakage Inside Containment."
- 4. NUREG-0609, "Asymmetric Blowdown Loads on PWR Primary Systems," 1981.
- 5. Generic Letter 84-04, "Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main Loops."
- 6. Letter from D. C. Dilanni, NRC, to R. W. Kober, RG&E,
Subject:
"Generic Letter 84-04," dated September 9, 1986.
- 7. NUREG-0821, "Integrated Plant Safety Assessment, Systematic Evaluation Program, R. E. Nuclear Power Plant," December 1982.
- 8. Letter from Guy S. Vissing (NRC) to Robert C. Mecredy (RG&E),
"Staff Review of the Submittal by Rochester Gas and Electric Company to Apply Leak-Before-Break Status to Portions of the R.E.
Ginna Nuclear Power Plant Residual Heat Removal System Piping", dated February 25, 1999.
R.E. Ginna Nuclear Power Plant B 3.4.15-6 Revision 62
RCS Specific Activity B 3.4.16 C.1 If the gross specific activity is not within limit, the change within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to MODE 3 and RCS average temperature < 500OF lowers the saturation pressure of the reactor coolant below the setpoints ofthe main steam safety valves and prevents automatically venting the SG to the environment in an SGTR event. The allowed Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 below 500OF from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.4.16.1 REQUIREMENTS This SR requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolant at least en .v,,y
.. 7 de*ys. While basically a quantitative measure of radionuclides with half lives longer than 15 minutes, excluding iodines, this measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in gross specific activity.
Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The Surveillance is applicable in MODES 1 and 2, and in MODE 3 with Tavg >_500OF. The 7 day Frtqu.n.y .. n.ide.S the unilkoliheed of a grooo fuel failuro duringM this timo.4\F.ý SR 3.4.16.2 This SR is only performed in MODE 1 to ensure iodine remains within limits during normal operation and following fast power changes when 3
7R fuel failure is more likely to occur. The 14 day Fru*n*-y is adequate t,
........... ., I .. The Frequency, between 2 and 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after a power change > 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results.
SR 3.4.16.3 A radiochemical analysis for E determination is required within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1
[INSERT 1 I operation have elaps since the reactor was last subcritical for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and evey , -- g (6 ,mnt,,) thereafter. This ensures that the radioactive materials are at equilibrium so the analysis for E is representative and not skewed by a crud burst or other similar abnormal event. The E determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit. The R.E. Ginna Nuclear Power Plant B 3.4.16-4 Revision 42
RCS Specific Activity B 3.4.16 analysis for E is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, JINSERT 3 excluding iodines . The i. - E .. ....
. .. u.....y Fc.g.. .... haigc.
- , d, V,,.
This SR is modified by a Note that indicates sampling is only required to be performed in MODE 1 such that equilibrium conditions are present during the sample.
REFERENCES 1. 10 CFR 50.67.
- 2. Design Analysis DA-NS-2001-084, Steam Generator Tube Rupture I Offsite and Control Room Doses.
R.E. Ginna Nuclear Power Plant B 3.4.16-5 Revision 42
Accumulators B 3.5.1 power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be faken to return the inoperable accumulator to OPERABLE status. The Completion Time minimizes the potential for exposure of the plant to a LOCA under these conditions.
The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowed to restore an inoperable accumulator to OPERABLE status is justified in WCAP-1 5049-A, Rev. 1 (Ref. 10).
C.1 and C.2 If the accumulator cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and pressurizer pressure reduced to
<_1600 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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.
D.1 If both accumulators are inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.5.1.1 REQUIREMENTS Each accumulator motor-operated isolation valve shall be verified to be fully open eveyr 12 het ,. Use of control board indication for valve position is an acceptable verification. This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve could result in not meeting accident analyses assumptions. Tie-F,,,*..u y is . .n.id. .dras.n. bl. M in ;vicw of ,the, administrativ.
that
.i1t .lIs a iiccd isolation unlikely..
e,,,uer va, c,isi,.c JINSERT 3P R.E. Ginna Nuclear Power Plant B 3.5.1-6 Revision 44
Accumulators B 3.5.1 SR 3.5.1.2 The borated water volume and nitrogen cover pressure shall be verified every 12 "he.... for each accumulator. This Fr..qucn.y is suffi... nt to nSur adequate i i injctiin during Ll.A. B n8 .f the stati. dUsigni
'eause ef the accumulator-, a 12 heur Froguoinoy usually allew th e~pcrater te identif; changcs b.f... lim^itS
" rc rcachcd. Main control b'ard alarm.s We else available fer th*s* a...umulater paramctorS. The level transmitters for the accumulators measure the level over a 14" span for the corresponding 0-100% level indicated on the main control board.
O~porating experienee h~as shown this Froqueney te be appropriate for eEarly d eteetien and eorrootioni ef off normal1tronds.
- ý SR 3.5.1.3 See SR 3.5.1.2 SR 3.5.1.4 The boron concentration shall be verified to be within required limits for each accumulator cvcr; 12 h9Ur3 by.. Me it...n inlcakegc. This is accomplished by monitoring the level ineach accumulator evefy 12 .het.s and comparing to the previous level readings. An unexplained increase in level could be an indication of inleakage and, therefore, dilution of the boron concentration. If an unexplained increase in level is detected, the ongoing change in boron concentration shall be determined by calculation. If the calculation indicates that the boron concentration had decreased to within 100 ppm of the lower limit, the affected accumulator shall be sampled to confirm boron concentration. In additin, accumulators shall be samgplcd eyer; 6 months to eeonfirmf that the borong conccentratien, infcrrcd frcm, inlakag t n limits.
Six mneiths is roaseinablo for Ycrifioation by sampling to dctcrmnine that eaeh aeoumnulator's borong eonccn~trationl i within the roquirod limgits, bcoausc the static design of the aocumulaIterS limits the ways Inwhiehthc-eonccntratien ean be changcd. This Frcqucnoy is adequate to identify ohanggos that could occur from mocehanismns, such as stratifloation or
~ifeakage.
SR3...
Verification eveiy-8-1-days that power is removed from each accumulator isolation valve operator when the pressurizer pressure is > 1600 psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, no accumulators would be available for injection if the LOCA were to occur in the cold leg containing the only OPERABLE accumulator. Siflee-pewet-is romovod undcr administrativc eontrol and Yalvo position is Ycrificd e-ve; 12 hoeurs, the 31 day Froguency will provide adequate asSUranee that power is rcmoved.
R.E. Ginna Nuclear Power Plant B 3.5.1-7 Revision 44
ECCS - MODES 1, 2, and 3 B 3.5.2 B.1 and B.2 If the inoperable train cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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.
C.1 If both trains of ECCS are inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be immediately entered. With one or more component(s) inoperable such that 100% of the flow equivalent to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analysis. Therefore, LCO 3.0.3 must be immediately entered.
SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Use of control board indication for valve position is an acceptable verification. Misalignment of these valves could render both ECCS trains inoperable. The listed valves are secured in position by removal of AC power or key locking the DC cortrol power. These valves are operated under administrative controls such that any changes with respect to the position of the valve breakers or key locks is unlikely. The verification of the valve breakers and key locks is performed by SR 3.5.2.3. Mispositioning of these valves can disable the function of both ECCS trains and invalidate the accident analyses. A heur Froequ Icis c~dercd rea3CnabC Vvinw eWf ethcr adMini~traltive ccntroks that enGurc a mtspesitiened valve i3 ulikety.
SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position. The 8 day Frcequcnc i3 ppciate beeause the valves arc epcraited under adminictratiye ecntrcl, and an improper valve pesitien in mcest eases, R.E. Ginna Nuclear Power Plant B 3.5.2-11 Revision 58
ECCS - MODES 1, 2, and 3 B 3.5..2 w.uld only affect a single train. This F..qu.n.y has been sh.wn t. be aeeoptabic through epcralting experienee. 4 LIINSEý[Jý SR 3.5.2.3 Verification e.v.ey..31 d*ys-that AC or DC power is removed, as appropriate, for each valve specified in SR 3.5.2.1 ensures that an active failure could not result in an undetected misposition of a valve which affects both trains of ECCS. Ifthis were to occur, no ECCS injection or recirculation would be available. Since power is romovo.d und.
admfinistrativo oontrol and valve position isYE~ified evor; 12 hourS, the 31 day Frequen^y will pro.vide adequate aSSUra.nc. that power* i 0 remv.MVed.
LiNSERT 3t SR 3.5.2.4 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component I problems is required by the ASME Code. This type of testing may be accomplished by measuring the pump developed head at a single point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis.
SRs are specified in the Inservice Testing Program, which encompasses I the ASME Code. The ASME Code provides the activities and Frequencies necessary to satisfy the requirements.
SR 3.5.2.5 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI signal and that each ECCS pump starts on receipt of an actual or simulated SI signal. This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 24 month Frequency is based on the need to prfr*mf thse Surf'eillanees under the conditions that apply during a plant cutagc and the potential for unplanned plant tralnsicnts if the Swr~eillanees werc performned with the reactor at power-. The 24 molnth Frequenoy is else aeceptable based en considcration of the design roliability (an conlfirmning epcrating cxperienee) of the eqjpmonet. The aetuation logic is tested as pa.. of ESF Aetuation System. testing, and equipmen performifanec is monitored as part of the lnseryiee Testing Programff.,
[INSERT 3 R.E. Ginna Nuclear Power Plant B 3.5.2-12 Revision 58
ECCS - MODES 1, 2, and 3 B 3.5.2 SR 3.5.2.6 See SR 3.5.2.5 SR 3.5.2.7 Periodic inspections of the containment sump suction inlet to the RHR System ensure that it is unrestricted and stays in proper operating condition. Th* 24 m^centh Frc..uen.y i. based en the need t. pc*, Fm this' Survc"Ilnec u,-v....,, ,,.... undcr n the
,.r..., ,,,,. eenditions that apply during irn~h-
,.v,,.,,,,,.VrtknThF-u r~ and a plant eutage, r. r to be 'uffie"^ntte dcteet abnormal dr, dntlpn and is 6.n*^Fed by
- pfitn
-,x..... enNSE T REFERENCES 1. Letter from R. A. Purple, NRC, to L. D. White, RG&E,
Subject:
"Issuance of Amendment 7 to Provisional Operating License No.
DPR-1 8," dated May 14, 1975.
- 2. Branch Technical Position (BTP) ICSB-1 8, "Application of the Single Failure Criterion to Manually-Controlled Electrically Operated Valves."
- 3. Letter from A. R. Johnson, NRC, to R. C. Mecredy, RG&E,
Subject:
"Issuance of Amendment No. 42 to Facility Operating License No.
DPR-18, R. E. Ginna Nuclear Power Plant (TAC No. 79829)," dated June 3, 1991.
- 4. Letter from D. M. Crutchfield, NRC, to J. E. Maier, RG&E,
Subject:
"SEP Topic VI-7.B: ESF Switchover from Injection to Recirculation Mode, Automatic ECCS Realignment, Ginna," dated December 31, 1981.
- 5. NUREG-0821.
- 6. UFSAR, Section 6.3.
- 7. Not Used
- 9. 10 CFR 50.46.
- 10. UFSAR, Section 15.6.
- 11. UFSAR, Section 6.2.
R.E. Ginna Nuclear Power Plant B 3.5.2-13 Revision 58
RWST B 3.5.4 ACTIONS A._1 With RWST boron concentration not within limits, it must be returned to within limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Under these conditions neither the ECCS nor the CS System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE condition. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> limit to restore the RWST boron concentration to within limits was developed considering the time required to change the boron concentration and the fact that the contents of the tank are still available for injection.
B. 1 With the RWST water volume not within limits, it must be restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In this Condition, neither the ECCS nor the CS System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE status or to place the plant in a MODE in which the RWST is not required. The short time limit of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to restore the RWST to OPERABLE status is based on this condition simultaneously affecting redundant trains.
C.1 and C.2 If the RWST cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.5.4.1 REQUIREMENTS The RWST water volume should be verified evey" 7,,,.. to be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and CS System pump operation on recirculation. Sin:^ the RWST^ vel-,. m .....
i* ,,Y stablc and the RWSTF is lccatcd in the Auxiiiar;y Building which prcViedcs suffieicnt Icak deteetien eapability, a :7 daY Froquonc is pppiate and has been shewn *.*:^ to be
- *-; I*.,,* acooptable 1.1*-,*- throlugh
,*, epefrating
'... -r*-,;; cxpeRionoo.
7 p.-*-*-; -
INSERTl 3 R.E. Ginna Nuclear Power Plant B 3.5.4-4 Revision 42
RWST B 3.5.4 SR 3.5.4.2 The boron concentration of the RWST should be verified evrey 74 .dys to be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA. Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized.
S*ien the RW"AIT vlu,, i" n,1.lly stable, a 7 day 1*apling F*rg*.n. y tc Ycrify beron conoontratiein ia 8ppropriato and has been shown to be acccptable through epcrating expcrienee.
REFERENCES 1. UFSAR, Section 3.11.
- 2. 10 CFR 50.49.
- 3. UFSAR, Section 6.3 and Chapter 15.
R.E. Ginna Nuclear Power Plant B 3.5.4-5 Revision 42
Containment Air Locks B 3.6.2 SR 3.6.2.2 The air lock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY. Thus, the door interlock feature supports containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur. -Due-te-the-purely mocehanical naturo of this interlock, and given that the interleeck mocehanism is only challenged when the containment airlock door i5 epencd, this test is enly rcgUircd tO be pe~fefmed onco every 24 moenths.
The 24 mnth Fr,*,," n.*'ny i. base ein g judgment and is
~nidered adequate in view of ethor indication -fdo aditel m~echanis~m status available to operaitiOns personol REFERENCES 1. UFSAR, Section 6.2.1.1.
- 2. 10 CFR 50, Appendix J, Option B.
R.E. Ginna Nuclear Power Plant B 3.6.2-7 Revision 21
Containment Isolation Boundaries B 3.6.3 E.2 Required Action E.2 requires that the mini-purge valve leakage must be restored to within limits, or the affected penetration flow path must be isolated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. 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 (including a single human error). For automatic valves, this requires two independent means to prevent the valve from re-opening.
Isolation barriers that meet this criterion are a closed and de-activated automatic valve, closed manual valve, or blind flange. A purge valve with resilient seals utilized to satisfy Required Action E.2 must have been demonstrated to meet the leakage requirements of SR 3.6.3.5. The specified Completion Time is reasonable, considering that one containment purge valve remains closed so that a major violation of containment does not exist.
Following completion of Required Action E. 1, verification that the affected penetration flow path remains isolated must be performed in accordance with Required Action D.2.
F.1 and F.2 If the Required Actions and associated Completion Times are not met, the plant must be brought to a MODE in which the LCO does not apply.
To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.6.3.1 REQUIREMENTS This SR ensures that the mini-purge valves are closed except when the valves are opened under administrative control. The mini-purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The valves may be opened for pressure control, ALARA or air quality considerations for personnel entry, maintenance activities, operational requirements, or for Surveillances that require the valves to be open. To be opened, the valves must be capable of closing under accident conditions, the containment isolation signal to the valves must be OPERABLE, and the effluent release must be monitored to ensure that it remains within regulatory limits. Th* 31 day Fr..qu.n.y is based en th^c.
rolative im~pertanee of these Yalyes sinec they provide a dircoet path to the-eutsfidc cnyergnment and the administrativc ccngtrols that are in plaec.
I IN S E*R*.T I3*,
I, R.E. Ginna Nuclear Power Plant B 3.6.3-11 Revision 64
Containment Isolation Boundaries B 3.6.3 SR 3.6.3.2 This SR requires verification that each containment isolation boundary located outside containment and not locked, sealed or otherwise secured in the required position is performing its containment isolation accident function. Containment isolation boundaries located beneath Appendix R fire wrap may be considered secured in the required position due to the administrative controls in place provided that a verification of the boundary position was made prior to securing the fire wrap. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment barrier is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown, that those isolation boundaries outside containment and capable of being mispositioned are in the correct position. This includes manual valves, blind flanges, pipe and end caps, and closed systems..ine. e.ntainmcnt is.lati. n b8un.a.io.
Wrc maintained under admfini~tratiye eentrols with eentainmmnt uselation beundary tags installed, the probability of their miselignment is low and a 92 day Frogucney to verif' thoir corroct pesition isappr.ria. ' t".e SR specifies that isolation boundaries that are open under admin=istra t N controls are not required to meet the SR during the time the boundarie) are open. JINET3 The SR is modified by two notes. The first Note applies to containment isolation boundaries located in high radiation areas and allows these boundaries to be verified closed by use of administrative means.
Allowing verification by administrative means (e.g., procedure control) is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3 and 4 for ALARA reasons. Therefore, the probability of misalignment of these isolation boundaries, once they have been verified to be in the proper position, is small. The Second Note states that this SR is not applicable to containment isolation boundaries which receive an automatic signal since the isolation times of these valves are verified by SR 3.6.3.4 and the boundaries are required to be OPERABLE.
SR 3.6.3.3 This SR requires verification that each containment isolation boundary located inside containment and not locked, sealed or otherwise secured in the required position is performing its containment isolation accident function. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment barrier is within design limits. This SR does not require any testing or valve manipulation.
Rather, it involves verification, through a system walkdown, that those isolation boundaries inside containment and capable of being mispositioned are in the correct position. This includes manual valves, blind flanges, pipe and end caps, and closed systems. Since containment isolation boundaries are maintained under administrative controls with containment isolation boundary tags installed, the R.E. Ginna Nuclear Power Plant B 3.6.3-12 Revision 64
Containment Isolation Boundaries B 3.6.3 SR 3.6.3.6 Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. Th* 24 m-nth Fr..u.ncy is based on the need to pci-formf this Sur~'cillanee under the conditions that apply during a plant eutagc and the pctential for an9 unplanned transient if the -Sur~'cillancc wcrc peffefficd with the roacter Bt pewcr. Operating cxpcricncc has shown that these compenents usually pass this Surveillanee when pecfeFmcd at the 24 mcn~th Frogueney. Thercfcrc, the Frcgucney was nudd to beacc.ptab. fr.m a liability standp.i..t.. INS.ERT 3J REFERENCES 1. Atomic Industry Forum GDC 53 and 57, issued for comment July 10, 1967.
- 2. Branch Technical Position CSB 6-4, "Containment Purging During Normal Operation."
- 3. UFSAR, Section 6.2.4 and Table 6.2-15.
- 4. Regulatory Guide 1.4, Revision 2.
- 5. 10 CFR 50, Appendix A, GDC 55, 56, and 57.
- 6. Ginna Station Procedure A-3.3.
- 7. NUREG-0800, Section 6.2.4.
R.E. Ginna Nuclear Power Plant B 3.6.3-14 Revision 64
Containment Pressure B 3.6.4 to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. However, due to the large containment free volume and limited size of the containment Mini-Purge System, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is allowed to restore containment pressure to within limits. This is justified by the low probability of a DBA during this time period.
B.1 and B.2 If containment pressure cannot be restored to within limits within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.6.4.1 REQUIREMENTS Verifying that containment pressure is within limits ensures that plant operation remains within the limits assumed in the containment analysis.
This verification should normally be performed using PI-944. The--, **heb*,
Frcquenc'; ef thus SR wasdevelped based en eperoting .xprinc rclt.d te trcnding. f containment prcaUr. vYrBitions durng th'^
applicable MO)DES. FwuthefRmcr, the 12 heur Froquenoy is considcrod-adequate min vie f ether indieatwions available lin the eentrol roomfi, 9including alarmis, to alert the opeffater to an9 abnormfal containmoent Calibration of PI-944 or other containment pressure monitoring devices should be performed in accordance with industry standards.
REFERENCES 1. UFSAR, Section 6.2.1.2.
R.E. Ginna Nuclear Power Plant B 3.6.4-3 Revision 72
Containment Air Temperature B 3.6.5 containment average air temperature within the limit is not required in MODE 5 or 6.
ACTIONS A.1 When containment average air temperature is not within the limit of the LCO, it must be restored to within the limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This Required Action is necessary to return operation to within the bounds of the containment analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is acceptable considering the sensitivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems.
B.1 and B.2 If the containment average air temperature cannot be restored to within its limit within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.6.5.1 REQUIREMENTS Verifying that containment average air temperature is within the LCO limit ensures that containment operation remains within the limit assumed for the containment analyses. In order to determine the containment average air temperature, an arithmetic average is calculated using measurements taken at locations within the containment selected to provide a representative sample of the overall containment atmosphere.
There are 6 containment air temperature indicators (TE-6031, TE-6035, TE-6036, TE-6037, TE-6038, and TE-6045) such that a minimum of three should be used for calculating the arithmetic average. The 4 2 he,*
Froquoney of this SR is eenoidorod aeooptablo based en ebserved slew Fates. f t,.mpr,,tu,. in...... within oontainm. nt as a. .sult
, f onvionmota! heat Sourocs (due to the lorgo velumo of eontainment).
FufthefRmor, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frogquonoy is eensiderod adequate inview ef ethor indioctiefns available in the ecntrcl rccm to alc~t the operater te an abnormI9al eentafinmont
^" .. .. . "^-'" . ... ... co .'.ilc
- teffpefraturoeni INSERT 3/
Calibration of these temperature indicators shall be performed in accordance with industry standards.
R.E. Ginna Nuclear Power Plant B 3.6.5-3 Revision 72
CS, CRFC and NaOH Systems B 3.6.6 D._l With one or two CRFC units inoperable, the inoperable CRFC unit(s) must be restored to OPERABLE status within 7 days. The inoperable CRFC units provided up to 100% of the containment heat removal needs.
The 7 day Completion Time is justified considering the redundant heat removal capabilities afforded by combinations of the CS System and CRFC System and the low probability of DBA occurring during this period.
E.1 and E.2 If the Required Action and associated Completion Time of Condition D of this LCO are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 With two CS trains inoperable, or three or more CRFC units inoperable, the plant is in a condition outside the accident analysis. Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.6.6.1 REQUIREMENTS The applicable SR descriptions from Bases 3.5.2 apply. This SR is required since the OPERABILITY of valves 896A and 896B is also required for the CS System.
SR 3.6.6.2 Verifying the correct alignment for manual, power operated, and automatic valves in the CS flow path provides assurance that the proper flow paths will exist for CS System operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. This SR does not require any testing or valve manipulation.
Rather, it involves verification, through a system walkdown, that those valves outside containment (there are no valves inside containment) and capable of potentially being mispositioned are in the correct position.
R.E. Ginna Nuclear Power Plant B 3.6.6-8 Revision 72
CS, CRFC and NaOH Systems B 3.6.6 SR 3.6.6.3 Verifying the correct alignment for manual, power operated, and automatic valves in the NaOH System flow path provides assurance that the proper flow paths will exist for NaOH System operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown, that those valves outside containment (there are no valves inside containment) and capable of potentially being mispositioned are in the correct position. ,*
SR 3.6.6.4 N Operating each CRFC unit for _ 15 minutes onco cvcry 31 days ensures that all CRFC units are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure, damper failures, or excessive vibration can be detected for corrective action. The A and C CRFC units must be operated with their respective charcoal filter train in the post accident alignment. The 31 day F..qu.n.y was devcleped considoring the knoewn Fliability of the fan units and eentrols, the rodundanoey available, and the low probability of significant dcgradation of the CRFC units occurring0 bcevccn survcillanees. It has elso been shown to be acccptable through opefrating expcriicnco.,
SIR 3.6.6.5 Verifying cooling water (i.e., SW) flow to each CRFC unit provides assurance that the energy removal capability of the CRFC assumed in the accident analyses will be achieved (Ref. 11). The minimum and maximum SW flows are not required to be specifically determined by this SR due to the potential for a containment air temperature transient.
Instead, this SR verifies that SW flow is available to each CRFC unit.
The 31 day Froguency was develeped congsidcring the known roliability of-the GW Systcmn, the twe CRFC train rcdundaney available, and the low probability of a significant degradlation of flew occurring bcevocn swefIfiaees-SIR 3.6.6.6 'INER3T Verifying each CS pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by the ASME Code (Ref. 12). Since the CS pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow.
This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice testing confirms component OPERABILITY, trends performance, and detects incipient failures by R.E. Ginna Nuclear Power Plant B 3.6.6-9 Revision 72
CS, CRFC and NaOH Systems B 3.6.6 abnormal performance. The Frequency of the SR is in accordance with the Inservice Testing Program.
SR 3.6.6.7 To provide effective iodine removal, the containment spray must be an alkaline solution. Since the RWST contents are normally acidic, the spray additive tank must provide a sufficient volume of spray additive to adjust pH for all water that is injected. This SR is performed to verify the availability of sufficient NaOH solution in the spray additive tank. Fhe484-day Frcqucncy was d...lpcd based.n the l.w , r... i-ity cf an undeteeted ehanggo in tank velume eeeurring during the SR interval 3incc the tanqk as ,.R,,lly 1 .Tank level is also indicated and alarmed in the control room, so that ther is high confidence that a substantial change in level would be dete ted.
SR 3.6.6.8 INSERT 3 This SR provides verification of the NaOH concentration in the spray additive tank and is sufficient to ensure that the spray solution being injected into containment is at the correct pH level. The 484 day Frogueney is suffiefient te encUrc that the eeneentratien lcvcl ef NeOH in the SPpraY additive tanik rcmain3 within the established lifnits. This as based on the lew likcliheed ef an uneentrclled ehag incccctratien sencc the tank ic nrm~fally iselated and the Prcbabii tha ay ubqtantial varianee *intank velumc will be detCeCe 4 \.
SR 3.6.6.9 This SR verifies that the required CRFC unit testing is performed in accordance with the VFTP. The VFTP includes testing HEPA filter performance. The minimum required flow rate through each of the four CRFC units is 33,000 cubic feet per minute at accident conditions (or 38,500 cubic feet per minute at normal operating conditions). Specific test frequencies and additional information are discussed in detail in the VFTP. However, the maximum surveillance interval for refueling outage tests is based on 24 month refueling cycles and not 18 month cycles as defined by Regulatory Guide 1.52 (Ref. 13).
SR 3.6.6.10 These SRs require verification that each automatic CS valve in the flowpath (860A and 860D) actuates to its correct position and that each CS pump starts upon receipt of an actual or simulated actuation of a containment High pressure signal. This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 24 moenth Froguoncey is based en the need to pcrferM these Safveillanees undor the eonditicns that the Survo"ieneop wero perfermed with the roaotfr at pewcr. Opefrating expericnee has shcwn that thesc eempenents usually pass thio R.E. Ginna Nuclear Power Plant B 3.6.6-10 Revision 72
CS, CRFC and NaOH Systems B 3.6.6 Survcillanecs when perfefrmcd at the 24 moenth F-r-qucney. Thcrofero, the Froequeney was ccncludcd to be acocptable frcmA a roliability-Stf,,.dp ei,.t. 11 d I SE T3 SR 3.6.6.11 ,INSERT3 See SR 3.6.6.10 SR 3.6.6.12 This SR requires verification that each CRFC unit, and the charcoal filter train associated with the A and C units, actuates upon receipt of an actual or simulated safety injection signal. The 24 month Fc..qu.n.y is based n igjudgment and has been shown to *b,,-heaeeptablo thr.ugh Eprtn cPcricncc. See SR 3.6.6.10 and SR 3.6.6.11, abeve-,6for further diseussion of the basis for the 24 mcnth Froequeney. ,
SR 3.6.6.13 This SR provides verification that each automatic valve in the NaOH System flow path that is not locked, sealed, or otherwise secured in position (836A and 836B) actuates to its correct position upon receipt of an actual or simulated actuation of a containment Hi-Hi pressure signal.
The 24 mo.nth frogu.ncy is based... oncginccrng judge.m.t and has been shewn to be aeecptable through oprtnexpeReiono. See SR 3.6.6.10 and 3.6.6.11, oSR abov, for futh'"r^ dis"-ussion of the baiS for the 24 mcenth Frequeonoy. ,
SR 3.6.6.14 To ensure that the correct pH level is established in the borated water solution provided by the CS System, flow through the eductor is verified
,onoc cvey 5 yceFs. This SR in conjunction with SR 3.6.6.13 provides JINSERT 1 "assurance that NaOH will be added into the flow path upon CS initiation.
A minimum flow of 20 gpm through the eductor must be established as assumed in the accident analyses. A flow path must also be verified from the NaOH tank to the eductors. Due to the passivna*t ur* of the SpFay additive flew controls, the 6 ycar Frouoc Is.ufficiont to identify
. mp*.nnt d....adation that mayf.. t flew injeetion.
SR 3.6.6.15 3 With the CS inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. As an alternative, a visual inspection (e.g. boroscope) of the nozzles or piping could be utilized in lieu of an air or smoke test if a visual inspection is determined to provide an equivalent or a more effective post-maintenance test. A visual inspection may be more effective if the potential for material intrusion is localized and the affected area is accessible. This SR ensures that each spray nozzle is unobstructed and provides assurance that spray coverage of the containment during an accident is not degraded. Due to the passive R.E. Ginna Nuclear Power Plant B 3.6.6-11 Revision 72
MSIVs and Non-Return Check Valves B 3.7.2 SR 3.7.2.3 This SR verifies that each MSIV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the plant to operation following a refueling outage. The MSIVs should not be tested at power, since even a partial stroke exercise increases the risk of a valve closure and plant transient when the plant is above MODE 4.
As the MSIVs are not tested at power, they are exempt from the ASME Code (Ref. 5), requirements during operation in MODES 1, 2 and 3.
The~ ffequejnew ef SR'V tes-toinn asrr ev94 Fflenths The 24 menth Frogueney fer testing is based en the rcfucling eyele. Gperatin cxperienee has shown that these eemponcnts usually pass the-Surveillamee when performoed att e-24 month Frcqu*n*l . Th*o*rfr, thi Reaueney is aeeemahle ffem a F-liability stardp3 REFERENCES 1. UFSAR, Section 5.4.4.
- 2. UFSAR, Section 15.1.5.
- 3. UFSAR, Section 3.6.2.5.1.
- 4. 10 CFR 50.67.
- 5. ASME Code for Operation and Maintenance of Nuclear Power I Plants.
R.E. Ginna Nuclear Power Plant B 3.7.2-6 Revision 58
ARVs B 3.7.4 SURVEILLANCE SR 3.7.4.1 REQUIREMENTS To perform a cooldown of the RCS, the ARVs must be able to be opened either remotely or locally. This SR ensures that the ARVs are tested through a full control cycle at least once per fuel cycle. Performance of inservice testing or use of an ARV during a plant cooldown may satisfy this requirement .... ting....expi'n" has sh*.w that these compenents usually pass the Suryeillanee whein pcrfermed at the 24 menth Froquenoy. The Froquoney is aeccptable fromn a roliabilt SR 3.7.4.2 The function of the block valve is to isolate a failed open ARV. Cycling the block valve both closed and open demonstrates its capability to perform this function. Performance of inservice testing or use of the block valve during plant cooldown may satisfy this requirement.
Operating cxpericnco has shewn that these eompenonts usually pass the Seu..illanoc when p...r.fo..d at the 24 mone^th F*o.u.n.y. The Frogquonoy is eeecptable Etrliability Yfrom standpoinA-REFERENCES 1. UFSAR, Section 10.3.2.5.
- 2. UFSAR, Section 15.6.3.
- 3. UFSAR, Section 15.1.6.
R.E. Ginna Nuclear Power Plant B 3.7.4-4 Revision 69
AFW System B 3.7.5 plant should not be perturbed by any action, including a power change, that might result in a trip. The seriousness of this condition requires that action be started immediately to restore one MDAFW, TDAFW, or SAFW train to OPERABLE status. For the purposes of this Required Action, only one TDAFW train flow path and the pump must be restored to exit this Condition.
Required Action H.1 is modified by a Note indicating that all required MODE changes or power reductions are suspended until one MDAFW, TDAFW, or SAFW train is restored to OPERABLE status. In this case, LCO 3.0.3 is not applicable because it could force the plant into a less safe condition.
SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the AFW and SAFW System water and steam supply flow paths provides assurance that the proper flow paths will exist for AFW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification, through a system walkdown, that those valves capable of being mispositioned are in the correct position.
The 31 day Frcqueney 09 based an cngnccn judgment, us eensistcnt with the procadural eeontrols gavarning pcain, Wevand ecnzUrcs corrcct valve pesitiens.
SR 3.7.5.2 3 Periodically comparing the reference differential pressure and flow of each AFW pump in accordance with the inservice testing requirements of the ASME Code (Ref. 4) detects trends that might be indicative of an incipient failure. The Frequency of this surveillance is specified in the Inservice Testing Program, which encompasses the ASME Code. The ASME Code provides the activities and Frequencies necessary to satisfy this requirement.
This SR is modified by a Note indicating that the SR is only required to be met prior to entering MODE 1 for the TDAFW pump since suitable test conditions have not been established. This deferral is required because there is insufficient steam pressure to perform the test.
R.E. Ginna Nuclear Power Plant B 3.7.5-8 Revision 66
AFW System B 3.7.5 SR 3.7.5.3 Periodically comparing the reference differential pressure and flow of each SAFW pump in accordance with the inservice testing requirements of the ASME Code (Ref. 4) detects trends that might be indicative of an incipient failure. Because it is undesirable to introduce SW into the SGs while they are operating, this testing is performed using the test condensate tank. The Frequency of this surveillance is specified in the Inservice Testing Program, which encompasses the ASME Code. The ASME Code provides the activities and Frequencies necessary to satisfy this requirement.
SR 3.7.5.4 This SR verifies that each AFW and SAFW motor operated suction valve from the SW System (4013, 4027, 4028, 9629A, and 9629B), each AFW and SAFW discharge motor operated valve (4007, 4008, 9701A, 9701B, 9704A, 9704B, and 9746), and each SAFW cross-tie motor operated valve (9703A and 9703B) can be operated when required. The Frequency of this Surveillance is specified in the Inservice Test Program and is consistent with the ASME Code (Ref. 4). The TDAFW discharge motor operated valve (3996) is maintained open and not required to be closed for the DBA's and transients described within the Applicable Safety Analyses section. Therefore, testing of the TDAFW discharge motor operating valve is not required.
SR 3.7.5.5 This SR verifies that AFW can be delivered to the appropriate SG in the event of any accident or transient that generates an actuation signal, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.
The 24 menth Froequonoy is based on the need to pcrferm thKi hSRveianSe undee the etnditions that apply durting a plant outage and the potontial for an unplanned transicnt if the Sur~'ci~ancc wor pcrfefrmod with the roaetcr at pewor. The 24 molnth Froequcnei aeeeptable based on eperating experienee and thec doi_ roiblity et the eettipmenle.
SR 3.7.5.6 -ISR3T This SR verifies that the AFW pumps will start in the event of any accident or transient that generates an actuation signal by demonstrating that each AFW pump starts automatically on an actual or simulated actuation signal. The 24 menth Fr..u.n.y is based en the po*tntial need to porforFM this Gurveillancc undor the conditions that apply during 8 plant eutage.
R.E. Ginna Nuclear Power Plant B 3.7.5-9 Revision 66
AFW System B 3.7.5 This SR is modified by a Note indicating that the SR is only required to be met prior to entering MODE 1 for the TDAFW pump since suitable test conditions may have not been established. This deferral is required because there is insufficient steam pressure to perform the test.
SR 3.7.5.7 This SR verifies that the SAFW System can be actuated and controlled from the control room. The SAFW System is assumed to be manually initiated within 14.5 minutes in the event that the preferred AFW System is inoperable. This Surveillance includes the verification of the automatic response of the motor operated discharge valves (9701 A and 9701 B) and the recirculation valves (9710A and 9710B). The F..*..uny cf 24 menths Is based en the need te perfeFRm thiS Suryeillanee under the ecnditions that apply dur*ing a plant eutagc and the petential fer ean unplannced tramseeint of the Survcfillancc wcrc pcrfcjfedat pewecj.
REFERENCES 1. UFSAR, Section 10.5.
- 2. UFSAR Chapter 15.
- 3. American National Standard, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," N18.2-1 973.
- 4. ASME Code for Operation and Maintenance of Nuclear Power Plants.
R.E. Ginna Nuclear Power Plant B 3.7.5-10 Revision 66
CSTs B 3.7.6 ACTIONS A.1 and A.2 If the CST water volume is not within limits, the OPERABILITY of the backup supply should be verified by administrative means within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
OPERABILITY of the backup feedwater supply must include verification that the flow paths from the backup water supply to the preferred AFW pumps are OPERABLE and immediately available upon AFW initiation, and that the backup supply has the required volume of water available.
Alternate sources of water include, but is not limited to, the SW System and the all-volatile-treatment condensate tank. In addition, the CSTs must be restored to OPERABLE status within 7 days, because the backup supply may be performing this function in addition to its normal functions. Continued verification of the backup supply is not required due to the large volume of water typically available from these alternate sources. The 7 day Completion Time is reasonable, based on an OPERABLE backup water supply being available, and the low probability of an event occurring during this time period requiring the CSTs.
B.1 and B.2 If the backup supply cannot be verified or the CSTs cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 wilhin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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 3.7.6.1 REQUIREMENTS This SR verifies that the CSTs contain the required volume of cooling water. The 24,350 gal minimum volume is met if one CST is >_22.8 ft (including instrument uncertainty) or if both CSTs are >_13.6 ft (including instrument uncertainty).. 12 b ^-
cxpericncc and the need fer epcr-ater awarcness ef plant evelutiens that
. . .- . ...hint... ............ 1q A ,Qq.,t== Q i F-e'uen^y :-nsiderTd is adequate in view 3f ethi t the1 eentrol room, ineluding alarms, to alert the epefrater te abnorma dcviations inthe CST-lvel-.
R.E. Ginna Nuclear Power Plant B 3.7.6-3 Revision 60
CCW System B 3.7.7 SURVEILLANCE SR 3.7.7.1 REQUIREMENTS Verifying the correct alignment for manual and power operated valves in the CCW flow path provides assurance that the proper flkw paths exist for CCW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification, through a system walkdown, that those valves capable of being mispositioned are in the correct position.
The 31 day F-r..u...y is based en ^ cngir crig judgment, as enss*". "nt with the p....dur.l c.ntr.ls gc,;cing ;'alvc . at'"n,
... and .. S... s eerroct valve pesitiens.
This SR is modified by a Note indicating that the isolation of the CCW flow to individual components may render those components inoperable but does not affect the OPERABILITY of the CCW loop header.
SR 3.7.7.2 This SR verifies that the two motor operated isolation valves to the RHR heat exchangers (738A and 738B) can be operated when required since the valves are normally maintained closed. The Frequency of this Surveillance is specified in the Inservice Test Program and is consistent with the ASME Code (Ref. 2).
REFERENCES 1. UFSAR, Section 9.2.2.
- 2. ASME Code for Operation and Maintenance of Nuclear Power I Plants.
R.E. Ginna Nuclear Power Plant B 3.7.7-6 Revision 58
SW System B 3.7.8 C.1 and C.2 If the SW pumps cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
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.
D.1 With three or more SW pumps or the loop header inoperable, the plant is in a condition outside of the accident analyses; therefore, LCO 3.0.3 must be entered immediately.
Required Action D.1 is modified by a Note requiring that the applicable Conditions and Required Actions of LCO 3.7.7, "CCW System," be entered for the component cooling water heat exchanger(s) made inoperable by SW. This note is provided since the inoperable SW system may prevent the plant from reaching MODE 5 as required by LCO 3.0.3 if both CCW heat exchangers are rendered inoperable.
SURVEILLANCE SR 3.7.8.1 REQUIREMENTS This SR verifies that adequate NPSH is available to operate the SW pumps and that the SW suction source temperature is within the limits assumed by the accident analyses and the system design. The-24-heL*-
Frcgqucncy iz based ein epcroting experienee rclated te tronding cf the parometcr YariotieHS dluring the applieeble MO)DES.
SR 3.7.8.2 3 Verifying the correct alignment for manual, power operated, and automatic valves in the SW flow path provides assurance that the proper flow paths exist for SW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to being locked, sealed, or secured. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation; rather, it involves verification, through a system walkdown, that those valves capable of being mispositioned are in the correct position.
R.E. Ginna Nuclear Power Plant B 3.7.8-7 Revision 68
SW System B 3.7.8 The 31 day F'*'.u.ny is based ,n gin .. i .. judgment,t. is n.i.t.nt with the procodural eeontrolS geyerningi vov oatien, and cnSUros corroct -vaiye positions.
This SR is modified by a Note indicating that the isolation of the SW flow to individual components or systems may render those components inoperable, but does not affect the OPERABILITY of the SW System.
SR 3.7.8.3 This SR verifies that all SW loop header cross-tie valves are locked in the correct position. This includes verification that manual valves 4623, 4639, 4640, 4665, 4668B, 4669, 4756, and 4760 are locked open and that manual valves 4610, 4611,4612, and 4779 are locked closed. The diesel generator cross-tie valves (4665, 4760, 4669, and 4668B) may be individually (one at a time) closed intermittently under administrative controls, such as during surveillance testing, as described in the LCO Bases. The 31 day F*r...u.n.y is based en _ngncring Judge'mnt, is c with the proi.dural eontr*osl go.o
.nsistent i cd o, and cnsuros corroct valve eDositlens fERTý3 NS SR 3.7.8.4 This SR verifies proper automatic operation of the SW motor operated isolation valves on an actual or simulated actuation signal (i.e., coincident safety injection and undervoltage signal). SW is a normally operating system that cannot be fully actuated as part of normal testing. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.
The 24 mointh Froqueney is based en the nooed tc pecfewm thi Tuhvnisvaneeundpr thoe anditions that apply duiong a plant putago and the portntial fae anunplanned tignsianthif inc l the ofro act psafetmd with the neartor at pswgn.OpSWating apinormall has shown that thcs eompenents usually pass the Surillano when poormaestingd at the 24 moenth Frog~uoncy. Thercfero, the Frcguceney is aepocpal fr~m. aM roliability standpoint. 4 SR 3.7.8.5 t-INE.RI3ýT This SR verifies proper automatic operation of the SW pumps on an actual or simulated actuation signal. This includes the actuation of the SW pumps following an undervoltage signal and following a coincident safety injection and undervoltage signal. SW is a normally operating system that cannot be fully actuated as part of normal testing during normal operation. The 24 moenth F-Fequcncy is based en the need to-p....... this
. ur.illane under the e"nditions that apply during a plant eutagc and the potential for anH unplanned transiont ifthe Sufveffillanee wc.. peio.....d with the roacter at power. GOpcating oexperiene has shown that these eempencnts usually pass the SurveJllanee when peofefrmed at the 24 moenth Frogqueney. Thcroforc, the Froequon-y i acooptablo fromH a roliability standpoint{.
R.E. Ginna Nuclear Power Plant B 3.7.8-8 Revision 68
CREATS B 3.7.9 a condition outside the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.7.9.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing each CREATS filtration train oncc cver,'31 days for _>15 minutes provides an adequate check of this system. The 31 day Frcgueney is based en the Feliabilit' ef the equipment, and the twe train rcdundancy .,.
SR 3.7.9.2 This SR verifies that the required CREATS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing the performance of the HEPA filter, charcoal adsorber efficiency, flow rate, and the physical properties of the activated charcoal. The required flowrate through each CREATS filtration train is 6000 cubic feet per minute (+/-10%). Specific test Frequencies and additional information are discussed in detail in the VFTP.
The value of 1.5% methyl iodide penetration was chosen for the laboratory test sample acceptance criteria because, even though the new system contains 4-inch charcoal beds, the design face velocity is 61 fpm.
Regulatory Guide 1.52, Revision 3 (Ref. 9), Table 1, provides testing criteria assuming a 40 fpm face velocity. The value of 1.5% was interpolated between the two values listed because of the higher face velocity of Ginna's system. The face velocity is listed in the specification because it is a non standard number. Testing at 61 fpm or greater satifies the criteria.
SR 3.7.9.3 This SR verifies that each CREATS train starts and operates and that each CREATS automatic damper actuates on an actual or simulated actuation signal. The Fr.quen.y.f 24 mon,.,th. is based en lnduIt.;-
op....ting cxpcricncc.
SR 3.7.9.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.
The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA R.E. Ginna Nuclear Power Plant B 3.7.9-6 Revision 51
ABVS B 3.7.10 ACTIONS A._1 When the ABVS is inoperable, action must be taken to place the plant in a condition in which the LCO does not apply. Action must be taken immediately to suspend movement of irradiated fuel assemblies in the Auxiliary Building. This does not preclude the movement of fuel to a safe position.
Required Action A.1 is modified by a Note indicating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies in the Auxiliary Building which have decayed < 60 days since being irradiated, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.
SURVEILLANCE SR 3.7.10.1 REQUIREMENTS This SR verifies the OPERABILITY of the ABVS. During fuel movement operations, the ABVS is designed to maintain a slight negative pressure in the Auxiliary Building to prevent unfiltered LEAKAGE. This SR ensures that Auxiliary Building exhaust fan C, and either Auxiliary Building main exhaust fan A or B are in operation and that the ABVS interlock mode switch is in the correct position. The Frcqu.ncy ef 24 heurS is based en enginccring judgcmcent and shewn te be aeccptable through cpefroting experienee. .
SR 3.7.10.2 "L--NSERT 3RT This SR verifies the integrity of the Auxiliary Building enclosure. The ability of the Auxiliary Building to maintain negative pressure with respect to the uncontaminated outside environment must be periodically verified to ensure proper functioning of the ABVS. During fuel movement operations, the ABVS is designed to maintain a slight negative pressure in the Auxiliary Building to prevent unfiltered leakage. This SR ensures that a negative pressure is being maintained in the Auxiliary Building.
The Frcequency ef 24 heura i5 based judgement and Fnognoing shewn te be eeeeptable through eporain oxprioo SR 3.7.10.3 This SR verifies that the required SFP Charcoal Adsorber System testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The SFP Charcoal Adsorber System filter tests are in general accordance with Regulatory Guide 1.52 (Ref. 5). The VFTP includes R.E. Ginna Nuclear Power Plant B 3.7.10-4 Revision 62
SFP Water Level B 3.7.11 ACTIONS A. 1 When the initial conditions assumed in the fuel handling accident analysis cannot be met, steps should be taken to preclude the accident from occurring. When the SFP water level is lower than tle required level, the movement of irradiated fuel assemblies in the SFP is immediately suspended. This action effectively precludes the occurrence of a fuel handling accident. This does not preclude movement of a fuel assembly to a safe position (e.g., movement to an available rack position).
Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply since if moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.
SURVEILLANCE SR 3.7.11.1 REQUIREMENTS This SR verifies sufficient SFP water is available in the event of a fuel handling accident. The water level in the spent fuel pool must be checked periodically during movement of irradiated fuel assemblies to ensure the fuel handling accident assumptions are met. The 7 day-Froquenoy is appropriate bcoa use the vo Iume in the peel iSnormaElly stable and the SFP is designed t3 provont drainago bclew 23 ft. Wator
!eyol changes arc controllcd by plant procodurco and arce eeooptflblc based ong eporating cxpericnco.
Verification of SEP water level can be accomplished by several means.
The top of the upper SEP pump suction line is 23 ft above the fuel stored in the pool. If there is Ž!23 ft of water above the reactor vessel flange (as required by LCO 3.9.6), with equal pressure in the containment and the Auxiliary Building, then at least 23 ft of water is available above the top of the active fuel in the storage racks.
In addition to the physical design features, there are two SEP level alarms (LAL 634) which are available to alert the operators of changing SEP level. A low level alarm will actuate when the SFP water level falls 4 inches or more from the normal level while a high level alarm will actuate when the SEP water level rises 4 inches or more from the normal level.
These alarms must receive a calibration consistent with industry practices before they are to be used to meet this SR.
R.E. Ginna Nuclear Power Plant B 3.7.11-3 Revision 59
SFP Boron Concentration B 3.7.12 APPLICABILITY This LCO applies whenever fuel assemblies are stored in the SFP to ensure the SFP keff remains
- 0.95 at all times.
ACTIONS A.1 and A.2 When the concentration of boron in the SFP is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.
This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The initiation of actions to restore concentration of boron is simultaneous with suspending movement of fuel assemblies.
The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply since if the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.
SURVEILLANCE SR 3.7.12.1 REQUIREMENTS This SR verifies that the concentration of boron in the SFP is within the limit. As long as this SR is met, the analyzed accidents are fully addressed. The 7 day Fr . i. appr"p.iate sin" e the b"-^ n is r with maintainingh.-,-*,-e F, subcr.itical. Alo,, the vlure and
- edited beren eeneentration in the peel is ncrrnally stable and all water level pfI eedul f Ve IIIs II.-III* Ll*** .l4J~s~l~ll -*l.vILI *J~ l*sI11 changes and bercn eeneentfatien changes are centrelled by plant R.E. Ginna Nuclear Power Plant B 3.7.12-3 Revision 20
Secondary Specific Activity B 3.7.14 ACTIONS A.1 and A.2 DOSE EQUIVALENT 1-131 exceeding the allowable value in the secondary coolant, is an indication of a problem in the RCS and contributes to increased post accident doses. If the secondary specific activity is not within limits the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.7.14.1 REQUIREMENTS This SR verifies that the secondary specific activity is within the limits of the accident analysis. A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT 1-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases. It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE. The 31 de, Fr..qu.n.y 0. based en the d.tcoti:n O*f*
incrcasing trcnd. of the 'eye' of DOSE EQUIVALENT 1'131, and all'w f8F apprcpriate actien to be talton to Fmaintain levels below the LCO) limnit.
.INSERT 31 REFERENCES 1. 10 CFR 50.67.
- 2. Design Analysis DA-NS-2002-007, Main Steam Line Break Offsite and Control Room Doses.
R.E. Ginna Nuclear Power Plant B 3.7.14-3 Revision 42
AC Sources - MODES 1, 2, 3, and 4 B 3.8.1 SURVEILLANCE The AC sources are designed to permit inspection and testing of all REQUIREMENTS important areas and features, especially those that have a standby function (Ref. 2). Periodic component tests are supplemented by extensive functional tests during refueling outages (under simulated accident conditions).
SR 3.8.1.1 This SR ensures proper circuit continuity for the independent offsite power source to each of the onsite 480 V safeguards buses and availability of offsite AC electrical power. Checking breaker alignment and indicated power availability verifies that each breaker is in its correct position to ensure that distribution buses and loads are connected to their qualified power source. The Fr, uen.y f*7- days ,- d,,,t, cincc' reaker p.siti.n is net likely to ch.nge withut thc pcratoros kn.wl.dgc and be"..use r.'m--- and indi"atficn
. f broakor arvailable in thc StuWe co,*trel rocFR.-
SR 3.8.1.2 This SR verifies that each DG starts from standby conditions and achieves rated voltage and frequency. This ensures the availability of the DG to mitigate DBAs and transients and to maintain the plant in a safe
.shutdown condition. The DG voltage control may be either in manual or automatic during the performance of this SR. The F...u.n.y of 31 days-is adequate to provide ass*u. nce of DG .PERABIIT.Y, whil. minmzing degradation reulting from. testing.
This SR is modified by two Notes. Note 1 indicates that performance of SR 3.8.1.9 satisfies this SR since SR 3.8.1.9 is a complete test of the DG.
The second Note states that all DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading. This minimizes the wear on moving parts that do not get lubricated when the engine is not running.
SR 3.8.1.3 This SR verifies that the DGs are capable of synchronizing with the offsite electrical system and accepting loads greater than or equal to the equivalent of the maximum expected accident loads. A minimum run time of 60 minutes is required to stabilize engine temperatures. A maximum run time of < 120 minutes minimizes the time that the DG is connected to the offsite source.
R.E. Ginna Nuclear Power Plant B 3.8.1-12 Revision 74
AC Sources - MODES 1, 2, 3, and 4 B 3.8.1 Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.85 lagging and 0.95 lagging. The upper load band limit of < 2250 kW is the DG two-hour rating and is provided to avoid routine overloading of the DG which may result in more frequent inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY. The lower band limit of 2025 kW bounds the maximum expected load following a DBA, based on worst case loading during the injection phase of the accident. The diesel generator loading will be below the long-term rating of 1950 kW within two hours.
In addition to verifying the DG capability for synchronizing with the offsite electrical system and accepting loads, the DG ventilation system should also be verified during this surveillance.
JINSERT 3 The Frcequeney of 31 days is adequate to proVido ac.Uranco of DG OPERABILITY, while minimizing degradetion rcsulting from tccting.
This SR is modified by four Notes. Note 1 indicates that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized. Note 2 states that momentary transients outside the load band (e.g., due to changing bus loads), do not invalidate this test. Similarly, momentary power factor transients above or below the administrative limit do not invalidate the test. Note 3 indicates that this Surveillance shall be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations. Note 4 stipulates a prerequisite requirement for performance of this SR. A successful performance of SR 3.8.1.2 or SR 3.8.1.9 must precede this surveillance to prevent unnecessary starts of the DGs.
S13-38. 1.4 This SR provides verification that the level of fuel oil in each day tank is at or above the minimum level, including instrument uncertainty, at which fuel oil is automatically added when the fuel oil transfer pump is in auto and the DG is operating. This level ensures adequate fuel oil for a minimum of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of DG operation at 110% of full load. A level of 8.75 inches, as read on the local sight glass, achieves these requirements.
TeFrcguency ef 31 days is adequato to ensurc that a sufficicnt supply of fuel ofil is available, sinee low lcycl alarmnS arc praVided and opcratoru SR 3.8.1.5 This SR demonstrates that each DG fuel oil transfer pump operates and transfers fuel oil from its associated storage tank to its associated day tank. This is required to support continuous operation of the DGs. This R.E. Ginna Nuclear Power Plant B 3.8.1-13 Revision 74
AC Sources - MODES 1, 2, 3, and 4 B 3.8.1 Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic or manual fuel transfer systems are OPERABLE.
The Frequency of 31 doys is ade quatc to provide assuranc of DG OPERABILITY, since the dcsign of thc fuel ill transfcr system is such that pumps .peratcautomatically o must bc sta.ted manually in-ordor to maintain an adequate vojume OT fuel om in the day tank flK9 cuinFo SR 3.8.1.6 This SR involves the transfer of the 480 V safeguards bus power supr ly from the 50/50 mode to the 100/0 mode and 0/100 mode which demonstrates the OPERABILITY of the alternate circuit distribution network to power the required loads. Thc Fr.qucnoy of 24 mo.nths i baseneig e dgment, taking into oonsidcration the plant
,.iti, nrquirca to pcncrm te
.urvciiiance, and is intcnaed to me consistent with cxpcct.dccylofuel lengths. Operating cxporionc has shown that thesc components usually pass the SR when porformoed at
=
iec -44 mnin rrcuucncv.1erfcr, 4
the.nrrU luency u was cunliuonu to aeceptagic tram a r eiiaaiiity standpoint.
SR 3.8.1.7 This SR verifies that each DG does not trip during and following a load rejection of > 295 kW. Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This SR demonstrates the DG load response characteristics and capability to reject the largest single load on the buses supplied by the DG (i.e., a safety injection pump).
In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, testing must be performed using a power factor _<0.9 lagging. This power factor is chosen to be representative of the actual design basis inductive loading that the DG would experience.
. -- =
The Freauc nov ot24 moenths into consideration plant een^i tiens r.quir*d to p the Su*veillance,
'rform and is intended to be cc INSERT, 3 This SR is modified by two Notes. The first Note states that this L- ......
Surveillance shall not be performed in MODE 1, 2, 3, or 4. The reason for the Note is that during operation in these MODES, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, plant R.E. Ginna Nuclear Power Plant B 3.8.1-14 Revision 74
AC Sources - MODES 1, 2, 3, and 4 B 3.8.1 safety systems. The second Note acknowledges that credit may be taken for unplanned events that satisfy this SR.
SR 3.8.1.8 This SR demonstrates that DG noncritical protective functions (e.g.,
overcurrent, reverse power, local stop pushbutton) are bypassed on an actual or simulated S! actuation signal. The noncritical trips are bypassed during DBAs but still provide an alarm on an abnormal engine condition. This alarm provides the operator with sufficient time to react appropriately. The DG availability to mitigate the DBA is more critical than protecting the engine against minor problems that are not immediately detrimental to emergency operation of the DG. The DG critical protective functions (engine overspeed, low lube oil pressure, and start failure (overcrank) relay) will be tested periodically per the station periodic maintenance program.
The Frzquency vf 24 ..... th. ia based "ngin*"en i"g judgment, taking iNto ecnsodortien plant cenditionB rcqUircd to pcrfcrm the Gufveillanoc, and is intended to be eensistcnt with cxpeotcd fuel cycle lcngh.
Operating cxpcricncc has shown that these compencnts usually pass the SR when perftrmed at th24-moenth FrFequency. T-herefore, this Frcqucney is aeceptablc fromn a reliability standpoit This SR is modified by two Notes. The first Note states that this Surveillance shall not be performed in MODE 1, 2, 3, or 4. The reason for the Note is that performing the Surveillance would remove a required DG from service which is undesirable in these MODES. The second Note acknowledges that credit may be taken for unplanned events that satisfy this SR.
SR 3.8.1.9 In the event of a DBA coincident with a loss of offsite power, the DGs are required to supply the necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded.
This SR demonstrates the DG operation during an actual or simulated loss of offsite power signal in conjunction with an actual or simulated SIl actuation signal. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.
Since it is not possible to operate all sequenced motors at their DBA loadings, a transient simulation program is used to demonstrate acceptable DG governor and voltage regulator operation. To successfully validate the testing data with the transient simulation program, the largest loads (with respect to both kW and current) must be sequenced on the R.E. Ginna Nuclear Power Plant B 3.8.1-15 Revision 74
AC Sources - MODES 1, 2, 3, and 4 B 3.8.1 DG during performance of this test. This includes two SI pumps, a CS IINSERT 3J- ,, \and RHR pump, and safety-related motor control centers, as a minimum.
The Frcquency of 24 moentha is based on cnginccrin~g judgement, taking inte considcratien plant conditiona9 rcguircd to performn the Surveillancc, and us intendled to be ecnsiotent with cxpccted fuel cyele lcngha This SR is modified by three Notes. Note 1 states that all DG starts may be preceded by an engine prelube period which is intended to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine lube oil continuously circulated and temperature maintained consistent with manufacturer recommendations for the DGs. Note 2 states that this Surveillance shall not be performed in MODE 1, 2, 3, or 4 since performing the Surveillance during these MODES would remove a required offsite circuit from service, cause perturbations to the electrical distribution systems, and challenge safety systems. Note 3 acknowledges that credit may be taken for unplanned events that satisfy this SR.
REFERENCES 1. UFSAR, Chapter 8.
- 3. UFSAR, Section 9.4.9.5.
- 4. UFSAR, Chapter 6.
- 5. UFSAR, Chapter 15.
- 7. "American National Standard, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," N18.2-1973.
- 8. Generic Letter 84-15, "Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability," July 2, 1984.
- 9. UFSAR Section 3.11 R.E. Ginna Nuclear Power Plant B 3.8.1-16 Revision 74
Diesel Fuel Oil B 3.8.3 time to correct high particulate levels prior to reaching the limit of acceptability. Poor sample practices (bottom sampling), contaminated sampling equipment, or errors in laboratory analysis can produce failures that do not follow a trend. Since the presence of particulates does not mean failure of the fuel oil to burn properly in the diesel engine, and particulate concentration is unlikely to change significantly between Surveillance Frequency intervals, and proper engine performance has been recently demonstrated (within 31 days), it is prudent to allow a brief period prior to declaring the associated DG inoperable. The 7 day Completion Time allows for further evaluation, resampling and re-analysis of the DG fuel oil.
C..1 With the new fuel oil properties defined in SR 3.8.3.2 not within required limits, a period of 30 days is allowed for restoring the stored fuel oil properties. This period provides sufficient time to test the stored fuel oil to determine that the new fuel oil, when mixed with previously stored fuel oil, remains acceptable, or to restore the stored fuel oil properties. This restoration may involve feed and bleed procedures, filtering, or combinations of these procedures. Even if a DG start and load was required during this time interval and the fuel oil properties were outside limits, there is a high likelihood that the DG would still be capable of performing its intended function.
D..1 With a Required Action and associated Completion Time not met, or one or more DG's fuel oil notwithin limits for reasons other than addressed by Conditions A, B, or C (e.g., cloud point temperature reached), the associated DG may be incapable of performing its intended function and must be immediately declared inoperable.
SURVEILLANCE SR 3.8.3.1 REQUIREMENTS This SR verifies an onsite supply of >_5000 gal of fuel oil is available for each required DG. This ensures that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> while providing maximum post-LOCA loads. The 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> period is sufficient time to place the plait in a safe shutdown condition and to bring in replenishment fuel from an offsite location.
The Frcqueney ef 31 days 09 adequate te ensuro that a sufficicnt supply ef fuel eel is available, s~incc indicaticnS arc available te einsur that epefrltefr Would be aWBro ef BAnY larg uses of fuel eil durin9g thiS pcriod.
R.E. Ginna Nuclear Power Plant B 3.8.3-3 Revision 48
DC Sources - MODES 1, 2, 3, and 4 B 3.8.4 SURVEILLANCE SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and tle ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery (or battery cell) and maintain the battery (or a battery cell) in a fully charged state. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations. The elevated equalize charge capability of the battery chargers is not an OPERABILITY requirement of the battery chargers and is not to be in service during the surveillance. The voltage drop when changing from the equalize conditions to the normal float conditions occurs relatively quickly. The 7 day F*euo*,
.. ny is nsistwnt with ma.enufacture.r r nmm-ndaticns and IEEEr450 (Rcf. 8)- ,
SR 3.8.4.2 3T This SR verifies that the capacity of each battery is adequate to supply and maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test. A battery service test is a special test of battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length corresponds to the design duty cycle requirements specified in Reference 2.
The Sur~oillanee Frogueney of 24 months is eensistcnt with the-rocommendatiens of Rogulater; Cuide 1.32 (Ref. 9) and Rcgulater;y Guide 1.129 (Ref. 10), whieh state that the battery zorwiee test should be pcrfermcd duarig rcfuelnlg oporaltiens or at seme other outagc, with intcrwais between, tests Het to exeeedd 24 moinths. t ISETý This SR is modified by two Notes. Note 1 states that SR 3.8.4.3 may be performed in lieu of SR 3.8.4.2. This substitution is acceptable because SR 3.8.4.3 represents a more severe test of battery capacity than does SR 3.8.4.2. Note 2 states that this surveillance shall not be performed in MODE 1, 2, 3, or 4 because performing the Surveillance would perturb the electrical distribution system and challenge safety systems.
SR 3.8.4.3 This Surveillance verifies that each battery capacity is Ž! 80% of the manufacturer's rating when subjected to a performance discharge test. A battery performance test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity as determined by specified acceptance R.E. Ginna Nuclear Power Plant B 3.8.4-6 Revision 41
DC Sources - MODES 1, 2, 3, and 4 B 3.8.4 criteria. The test is intended to determine overall battery degradation due to age and usage.
A battery should be replaced if its capacity is below 80% of the manufacturer rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements. ,-,, I.NERT 1 The Frequency for this SR is 6A menths when the battery is < 85% of its expected life with no degradation and 12 months if the battery shows degradation or has reached 85% of its expected life with a capacity
< 100% of the manufacturer's rating. When the battery has reached 85%
of its expected life with capacity Ž_100% of the manufacturer's rating, the Frequency becomes 24 months. Battery degradation is indicated when the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is > 10% below the manufacturer rating. These Frequencies are considered acceptable based on the testing being performed in a conservative manner relative to the battery life and degradation. This ensures that battery capacity is adequately monitored and that the battery remains capable of performing its intended function.
This SR is modified by a Note stating that this SR shall not be performed in MODE 1, 2, 3, or 4. The reason for the Note is that during operation in these MODES, performance of this SR could cause perturbations to the electrical distribution system and challenge safety systems.
R.E. Ginna Nuclear Power Plant B 3.8.4-7 Revision 41
DC Sources - MODES 1, 2, 3, and 4 B 3.8.4 July
- 1. Atomic Industrial Forum (AIF) GDC 39, Issued for comment REFERENCES 1. Atomic Industrial Forum (AIF) GDC 39, Issued for comment July 10,1967.
- 2. UFSAR, Section 8.3.2.
- 3. UFSAR, Section 9.4.9.3.
- 4. UFSAR, Chapter 6.
- 5. UFSAR, Chapter 15.
- 6. UFSAR, Section 8.3.1.
- 8. IEEE-450-1980.
__,,1/eleeleted
- 9. R.gul.t'r-y Guide 1.32, Fcbruor; 7...'
- 10. R.gul.ety Cuide 1..129, Dc......r 1974.
R.E. Ginna Nuclear Power Plant B 3.8.4-8 Revision 41
Battery Cell Parameters B 3.8.6 SURVEILLANCE SR 3.8.6.1 REQUIREMENTS This SR verifies that the electrolyte level of each connected battery cell is above the top of the plates and not overflowing. This is consistent with IEEE-450 (Ref. 4) and ensures that the plates suffer no physical damage and maintain adequate electron transfer capability. The F,-,qucn.y of 31 days is eensistcnt with IEEE 450.
SR 3.8.6.2 3 This SR verifies that the float voltage of each connected battery cell is
> 2.07 V. This limit is based on IEEE-450 (Ref. 4) which slates that a cell voltage of 2.07 V or below, under float conditions and not caused by elevated temperature of the cell, indicates internal cell problems and may require cell replacement. Th* f*r..u.ny of 31 days is else.. nsistcnt with IEEE 450.
SR 3.8.6.3 -iINSERT 3T This SR verifies the specific gravity of the designated pilot cell in each battery is _>1.195. This value is based on manufacturer recommendations. According to IEEE-450 (Ref. 4), the specific gravity readings are based on a temperature of 77°F (251C). The specific gravity readings are corrected for actual electrolyte temperature. For each 30 F (1.671C) above 77 0 F (25 0 C), 1 point (0.001) is added to the reading; 1 point is subtracted for each 3 0 F below 77°F. The specific gravity of the electrolyte in a cell increases with a loss of water due to electrolysis or evaporation.
Because of specific gravity gradients that are produced during the recharging process, delays of several days may occur while waiting for the specific gravity to stabilize. A stabilized charger current is an acceptable alternative to specific gravity measurement for determining the state of charge. This phenomenon is further discussed in IEEE-450.
The Frogueney ef 31 days as eensistent with IEEE 450.
SR 3.8.6.4 INERT 3 This SR verifies the average electrolyte temperature of the designated pilot cell in each battery is > 55 0 F. This temperature limit is an initial assumption of the battery capacity calculations. The Fr.qu.n.y .f 31 days is eensistent with IEEE 450 (Ref-. 4.
R.E. Ginna Nuclear Power Plant B 3.8.6-3 Revision 40
Battery Cell Parameters B 3.8.6 SR 3.8.6.5 This SR verifies that the average temperature of every fifth cell of each battery is > 550F. This is consistent with the recommendations of IEEE-450 (Ref. 4). Lower than normal temperatures act to inhibit or reduce battery capacity. This SR ensures that the operating temperatures remain within an acceptable operating range. The FFr.u.ncy.f 92 days-0s ccnsistent with IEEE 450.
SR 3.8.6.6 3 This SR verifies the specific gravity of each connected cell is not more than 0.020 below average of all connected cells and that the average of all connected cells is > 1.195. This value is based on manufacturer recommendations and IEEE-450 (Ref. 4) which ensures that the effect of a highly charged or new cell does not mask overall degradation of the battery. The temperature correction for specific gravity readings is the same as that discussed for SR 3.8.6.3. The Fr..u.n.y ef 92 del. iC ccnsste.nt with I-EEE 450-REFERENCES 1. UFSAR, Section 3.8.2.
- 2. UFSAR, Chapter 6.
- 3. UFSAR, Chapter 15.
- 4. IEEE-450-1980.
R.E. Ginna Nuclear Power Plant B 3.8.6-4 Revision 40
AC Instrument Bus Sources - MODES 1, 2, 3, and 4 B 3.8.7 SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This SR verifies correct static switch alignment to Instrument Bus A and C. This verifies that the inverters are functioning properly and AC Instrument Bus Aand C are energized from their respective inverter. The verification ensures that the required power is available for the instrumentation of the RPS and ESF connected to the AC instrument buses. The F..qucn.y f.7 days takes int' ae-e"nt the,-dundant capability ef the invclter.
.thc1 and indi tinsI ayailabic inI the lt*IlI rcamn that alert the epeffltfr te inverter mfalfuncticems.~
SR 3.8.7.2 This SR verifies the correct Class 1E CVT alignment to Instrument Bus B.
This verifies that the Class 1 E CVT is functioning properly and supplying power to AC Instrument Bus B. The verification ensures that the requred power is available for the instrumentation of the RPS and ESF connected to the AC instrument bus. The F.qu.n.y .f 7- days takes int, a.. .unt the rdundint i'*iAi Umnt buses and ther ind.iaticnsea*Vilable in the ccntrcl rccm that alert the epefrater to the Glass 1E CYT mc~lfuctc REFERENCES 1. UFSAR, Chapter 8.3.2.
- 2. UFSAR, Chapter 6.
- 3. UFSAR, Chapter 15.
- 4. UFSAR, Section 8.3.1.
R.E. Ginna Nuclear Power Plant B 3.8.7-6 Revision 41
AC Instrument Bus Sources - MODES 5 and 6 B 3.8.8 The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required AC instrument bus power source should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power or powered from an alternate power source.
SURVEILLANCE SR 3.8.8.1 REQUIREMENTS This SR verifies correct static switch alignment to the required AC instrument buses. This SR verifies that the inverter is functioning properly and the AC instrument bus is energized from the inverter. The verification ensures that the required power is available for the instrumentation connected to the AC instrument bus. The F.rqu.n.ycf 7 days takes inte aeeeunt the rodundant capability of the finycrtcr and ether indfieaticns available in the control rocm that alert the epefrater to inverter SR 3.8.8.2 3 This SR verifies the correct Class 1E CVT alignment when Instrument Bus B is required. This verifies that the Class 1E CVT is functioning properly and supplying power to AC Instrument Bus B. The verification ensures that the required power is available for the instrumentation of the RPS and ESF connected to the AC instrument bus. The Fr.qu.ncy cf 7 days takes into aeecwunt the rcdundant inotrument buses and ethcr indieations available in the contrOl room that alert the epefrater to the Class! Er- Cl Falfuneticc,-A:^
REFERENCES 1. None.
R.E. Ginna Nuclear Power Plant B 3.8.8-5 Revision 61
Distribution Systems - MODES 1, 2, 3, and 4 B 3.8.9 D.1 and D.2 If the inoperable distribution subsystem cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 With two trains with inoperable electrical power distribution subsystems, the potential for a loss of safety function is greater. If a loss of safety function exists, no additional time is justified for continued operation and LCO 3.0.3 must be entered. This Condition may be entered with the loss of two trains of the same electrical power distribution subsystem, or with loss of Train A of one electrical power distribution subsystem coincident with the loss of Train B of a second electrical power distribution subsystem such that a loss of safety function exists.
SURVEILLANCE SR 3.8.9.1 REQUIREMENTS This SR verifies that the electrical power trains are functioning properly, with all required power source circuit breakers closed, tie-breakers open, and the buses energized from their allowable power sources. Required voltage for the AC electrical power distribution subsystem is _ 420 VAC; for the DC electrical power distribution subsystem, Ž_108.6 VDC and
< 140 VDC; and for AC instrument bus electrical power distribution subsystem, between 113 VAC and 123 VAC at the instrument buses.
Required voltage for the instrument distribution panels is between 110 VAC and 123 VAC. Required voltage for inverter MQ-483 is between 107 volts and 129.8 volts. The loss of inverter MQ-483 is addressed in LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)
Instrumentation" and LCO 3.3.3, "Post Accident Monitoring (PAM)
Instrumentation" for the affected individual containment wide range pressure and steam generator B pressure instrumentation (PT-950 and PT-479). The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. T-he-F...u.n.y of 7 days t*kes int" account the "edundent
.apability
' f the AC, D, n.. ACG,lctr..al nO.trumnt bus pewe. di.t.ibu.tien
- ub-systcm, and ether in~dicatiens available inthe control reoom that alcrt the eperater to stubsystcm malfunctions.
R.E. Ginna Nuclear Power Plant B 3.8.9-9 Revision 68
Distribution Systems - MODES 5 and 6 B 3.8.10 Therefore, Required Action A.2.5 requires declaring RHR inoperable, which results in taking the appropriate RHR actions.
The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required distribution subsystems should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power.
SURVEILLANCE SR 3.8.10.1 REQUIREMENTS This Surveillance verifies that the electrical power distribution trains are functioning properly, with all the required power source circuit breakers closed, required tie-breakers open, and the required buses energized from their allowable power sources. Required voltage for the AC power distribution electrical subsystem is >_420 VAC, for the DC power distribution electrical subsystem > 108.6 VDC and _<140 VDC, and for AC instrument bus power distribution electrical subsystem is between 113 VAC and 123 VAC at the instrument buses. Required voltage for the instrument distribution panels is between 110 VAC and 123 VAC.
Required voltage for inverter MQ-483 is between 107 volts and 129.8 volts. The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. Trhe-Frcgqucrny ef 7 days takes into account the eapability of the AC, DCG, and AC instrumonet bus eleetrical pewer diStributien subsystems, and ether inmdications available inthe control roomf that alcrt the epcratert subsystem mtalfunctions-.
REFERENCES 1. None.
R.E. Ginna Nuclear Power Plant B 3.8.10-6 Revision 61
Boron Concentration B 3.9.1 There are no safety analysis assumptions of boration flow rate and concentration that must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible. In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for plant conditions.
Once action has been initiated, it must be continued until the boron concentration is restored. The restoration time deperds on the amount of boron that must be injected to reach the required concentration.
SURVEILLANCE SR 3.9.1.1 REQUIREMENTS This SR ensures the coolant boron concentration of the refueling canal, the refueling cavity, and the portions of the RCS that are hydraulically coupled, is within the COLR limits. The boron concentration of the coolant is determined by chemical analysis. The sample should be representative of the portions of the RCS, the refueling canal, and the refueling cavity that are hydraulically coupled with the reactor core.
A FFeueieyefene vecr;7 heLurS iS Faseel e metunt ef timne Yerify the bercn eenecntratien ef the rcprcsentative sample(s). The te Ibe l d.q ued.lI y V -~i.'q J 1*J,,1.v*l* l* 'i,,iq*I *nq4.qe ,ll*J,,llI ,llilq.
Frcquency is based on epcroiting cxpcrienee, whieh has shewn :72 hours REFERENCES 1. Atomic Industrial Forum (AIF) GDC 27, Issued for comment July 10, 1967.
- 2. UFSAR, Section 15.4.4.2.
- 3. NUREG-0800, Section 15.4.6.
R.E. Ginna Nuclear Power Plant B 3.9.1-4 Revision 61
Nuclear Instrumentation B 3.9.2 The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is sufficient to obtain and analyze coolant samples for boron concentration. The Frequency of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensures unplanned changes in boron concentration would be identified. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable, considering the low probability of a change in core reactivity during this time period.
C.1, C.2, and C.3 With no audible count rate available, only visual indication is available and prompt and definite indication of a boron dilution event has been lcst.
Therefore, CORE ALTERATIONS and positive reactivity additions must be suspended immediately. Performance of Required Actions C.1 and C.2 shall not preclude completion of movement of a component to a safe position (i.e., other than a normal cooldown of the coolant volume for the purpose of system temperature control within established procedures).
Since CORE ALTERATIONS and positive reactivity additions are not to be made, the core reactivity condition is stabilized until the audible count rate capability is restored. This stabilized condition is determined by performing SR 3.9.1.1 to ensure that the required boron concentration exists.
The Completion time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is sufficient to obtain and analyze coolant samples for boron concentration. The Frequency of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensures unplanned changes in boron concentration would be identified.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable, considering the low probability of a change in core reactivity during this time period.
SURVEILLANCE SR 3.9.2.1 REQUIREMENTS This SR is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one monitor to a similar parameter on another monitor. It is based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading and core geometry can result in significant differences between source range monitors, but each monitor should be consistent with its local conditions.
The inoperability of one source range neutron flux channel prevents performance of a CHANNEL CHECK for the operable channel. However, the Required Actions for the inoperable channel requires suspension of CORE ALTERATIONS and positive reactivity addition such that the CHANNEL CHECK of the operable channel can consist of ensuring consistency with known core conditions.
R.E. Ginna Nuclear Power Plant B 3.9.2-3 Revision 61
.Nuclear Instrumentation B 3.9.2 The Frcgueney 3f 12 heurs is eensistent with the CHANNEL G CHECK IIFrcuenly spc.ificd simIlrly for thc Ioll iilFthelsate ii C -3.3.4,
'Reaeter TrFip System (RTSG) Instrumentatien."
SR 3.9.2.2 This SR is the performance of a CHANNEL CALIBRATION every24-tenlnhs. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. The CHANNEL CALIBRATION for the source range neutron flux monitors consists of obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to baseline data. The 24 m-nth FroguenLy_urv1illanee is based en the need te perform this undrr the eenditiens that apply durfing a plant outage. O~perating experienee has shown these eempenents usually pass the Survoillanco when perfefrmod Rim1 46 mFRRR3 TFRl -. I- Fequefiy REFERENCES 1. UFSAR, Section 7.7.3.2.
RE. Ginna Nuclear Power Plant B 3.9.2-4 Revision 61
Containment Penetrations B 3.9.3 ACTIONS A.1 and A.2 If the containment equipment hatch (or its closure plate or roll up door and associated enclosure building), air lock doors, or any containment penetration that provides direct access fromthe containment atmosphere to the outside atmosphere is not in the required status, including the Containment Ventilation Isolation System not capable of automatic actuation when the purge and exhaust valves are open, the plant must be placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This SR demonstrates that each of the containment penetrations are in the required status. The Surveillance on the open purge and exhaust valves will demonstrate that the valves are not blocked or otherwise prevented from closing (e.g., solenoid unable to vent).
The SurfvIe;n* is ooe.Formd eve.y 7 days during CORE
......... L OA J AL 1LLjAI N*ISI of e4 odiatcd
.neveffcnt **,ul assemblies win e-ntainent. The Surmillan;e . is "craIselected t3 be eemmensurtc with the nirmal durati n ,ffi.fl tc eeffmpl*t* fuel handling ep.* atic;s. As such, this Surveillanee cn9Surcs that a postulated fuel handling eeeident that rolcases fission product radioactivity within the eontainmcnt will noet
.... i I ,sUiltin a *.luas e,f fissiun produ.t radlieaotlvity to tr eccnVIronment.
SR 3.9.3.2 JINSERT3 This SR demonstrates that each containment purge and exhaust valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal. The 24 month Fro......y mainta::ins unsistny with othnr snim ilar .. ntati.
.... and v
.inSt t .lv. .stig roguiromolnts. In LCO)3.3.5, the Containmcnt Vcntlation Isolatio instrumentatien3.roguirca a CHANNEL I II,... Jl l*..l
.ll '**~VSl £* I.,V
- Ic
- 'J ,L ll% v,J CHECK cvcr; 24 heurs an~d a COT:
V%# l*.l l lV.l k y .l kl.ll,* l l l.tl l ever; 92 days to cnsurc the chanrncl O)PERABILITY during r-efucling
ý ý--! Mvý 'A ý 4ký^ At-r! WArlkK! ! f'Ir' =0'r -A CHANNE6
- AL*BRATION is pIFrfeord. These S.eIrilla*.. s will ensU :'aT-p that the valves arc capablc of closing after a postulated fuel handling eeewdent to limfit 8 rolcase of fission product radioactivity fromg the R.E. Ginna Nuclear Power Plant B 3.9.3-4 Revision 53
RHR and Coolant Circulation - Water Level 2! 23 Ft B 3.9.4 A.4 If RHR loop requirements are not met, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. With the RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded.
The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on the low probability of the coolant boiling in that time.
SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This SR requires verification e;e'y 12 heotr that one RHR loop is in operation and circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providig decay heat removal capability and mixing of the borated coolant to prevent thermal and boron stratification in the core. The-Frcqueney
.......... ef 12 te
,-8FH ....
the epf.e heUFS 69 suffienent th..*^.*
"e "ete eensidering indmeatiens
,eeethcrte'^Fne,4:^
- *H and e, p REFERENCES 1. UFSAR, Section 5.4.5.
- 2. UFSAR, Section 15.4.4.2.
R.E. Ginna Nuclear Power Plant B 3.9.4-4 Revision 61
RHR and Coolant Circulation - Water Level < 23 Ft B 3.9.5 SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This SR requires verification evety 12 hlret, that one RHR loop is in operation and circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing decay heat removal capability and mixing of the borated coolant to prevent thermal and boron stratification in the core. T-he-Frcequeney of 12 heurs is suffricgnt considcring ether indicatiens and alarmS available tc the eporatr in the eon*trl ro;m tc m.nite- RHR l*p*
pe~fefffle~ee SR 3.9.5.2 I Verification that a second RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the standby pump. The F...qucn.y of 7 days is. r.as.nabl. in Yi.w
.nsidorcd .f ether adFmignstrativc controls available and has been shown tob ooeptableby operating oxporion~oo.
REFERENCES 1. UFSAR, Section 5.4.5.
- 2. UFSAR, Section 15.4.4.
R.E. Ginna Nuclear Power Plant B 3.9.5-4 Revision 61
Refueling Cavity Water Level B 3.9.6 LCO A minimum refueling cavity water level of 23 ft above the reactor vessel flange is required to ensure the radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits and preserves the assumptions of the fuel handling accident analysis (Ref. 1). As such, it is the minimum required level during movement of fuel assemblies wthin containment. Maintaining this minimum water level in the refueling cavity also ensures that > 23 ft of water is available in the spent fuel pool during fuel movement assuming that containment and Auxiliary Building atmospheric pressures are equal.
APPLICABILITY This LCO is applicable when moving irradiated fuel assemblies within containment. This LCO is also applicable during CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts. The LCO ensures a sufficient level of water is present in therefueling cavity to minimize the radiological consequences of a fuel handling accident in containment. Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3.7.11, "Spent Fuel Pool (SFP) Water Level."
ACTIONS A.1 and A.2 When the initial condition assumed in the fuel handling accident cannot be met, steps should be taken to preclude the accident from occurring.
With a water level of < 23 ft above the top of the reactor vessel flange, all operations involving CORE ALTERATIONS or movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur.
The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum refueling cavity water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met. VAter at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 1).
R.E. Ginna Nuclear Power Plant B 3.9.6-2 Revision 59
Refueling Cavity Water Level B 3.9.6 The F-rc.ucn.y 24
.f hurs is bascd cn coing judgment
.a and eemsidered adequate inview ef the large Yealuin ef water and the neFrrnal p....du.al
... ntr'sl'f valve p..ition:, which m9o" :ignifi-ant unplan'..d level ehanges unlikely.
REFERENCES 1. UFSAR, Section 15.7.3.
- 2. 10 CFR 50.67.
R.E. Ginna Nuclear Power Plant B 3.9.6-3 Revision 59
ATTACHMENT 5 License Amendment Request R. E. Ginna Nuclear Power Plant Docket No. 50-244 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
TSTF-425 (NUREG-1431) vs. Ginna Cross-Reference
LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-244 Page 1 of 10 TSTF-425 (NUREG-1431) vs. Ginna Cross-Reference Technical Specification Section Title/Surveillance Description* TSTF-425 GINNA Definitions 1.1 1.1 Staggered Test Testing Shutdown margin (SDM) 3.1.1 3.1.1 Verify SDM within limits 3.1.1.1 3.1.1.1 Core Reactivity 3.1.2 3.1.2 Verify core reactivity within predicted value 3.1.2.1 3.1.2.2 Rod Group Alignment Limits 3.1.4 3.1.4 Verify individual rod position within alignment 3.1.4.1 3.1.4.1 Verify rod freedom of movement (trippability) 3.1.4.2 3.1.4.3 Verify individual rod position within alignment with rod position 3.1.4.2 monitor inoperable Shutdown Bank Insertion Limits 3.1.5 3.1.5 Verify shutdown bank within insertion limit per COLR 3.1.5.1 3.1.5.1 Control Bank Insertion Limit 3.1.6 3.1.6 Verify control bank within insertion limit per COLR 3.1.6.2 3.1.6.2 Verify sequence and overlap limits per COLR 3.1.6.3 3.1.6.4 Verify control bank within insertion limit per COLR when insertion 3.1.6.3 limit monitor inoperable Physics Tests (Exceptions Mode 2) 3.1.8 3.1.8 Verify RCS lowest loop temperature 3.1.8.2 3.1.8.2 Verify thermal power 3.1.8.3 3.1.8.3 Verify SDM within COLR 3.1.8.4 3.1.8.4 Fq(Z) 3.2.1 Verify measured Fq(Z) 3.2.1.1 Verify measured FWxV <Fx' 3.2.1.2 F,(Z) (RAOC-W(Z) Methodology 3.2.1B 3.2.1 Verify F c%(Z) is within limit 3.2.1.1 3.2.1.1 Verify F Wq(Z) is within limit 3.2.1.2 3.2.1.2 Fq(Z) (CAOC-W(Z) Methodology 3.2.1C Verify F cq(Z) is within limit 3.2.1.1 Verify F Wq(Z) is within limit 3.2.1.2 Nuclear Enthalpy Rise Hot Channel Factor (FN dh) 3.2.2 3.2.2 Verify Fdh within limits per COLR 3.2.2.1 3.2.2.1 Verify Fdh within limits per COLR (only when one power range 3.2.2.2 inoperable)
AFD 3.2.3NB Verify AFD is within limits 3.2.3.1 Update target flux difference 3.2.3.2 Determine by measurement target flux difference 3.2.3.3 AFD(RAOC Methodology) 3.2.3B 3.2.3 Verify AFD is within limits 3.2.3.1 3.2.3.1 QPTR 3.2.4 3.2.4 Verify QPTR within limits by calculation 3.2.4.1 3.2.4.1 Verify QPTR within limits by incore detectors 3.2.4.2 3.2.4.2
LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-244 Page 2 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 GINNA RTS Instrumentation 3.3.1 3.3.1 Channel Check 3.3.1.1 3.3.1.1 Compare calorimetric to power range channel output 3.3.1.2 3.3.1.2 Compare results of incore to NIS 3.3.1.3 3.3.1.3 Perform. TADOT 3.3.1.4 3.3.1.4 Perform ACTUATION LOGIC TEST 3.3.1.5 3.3.1.5 Calibrate excore channels to agree with incore 3.3.1.6 3.3.1.6 Perform COT 3.3.1.7 3.3.1.7 Perform COT 3.3.1.8 3.3.1.8 Perform TADOT 3.3.1.9 3.3.1.9 Channel Calibration 3.3.1.10 Channel Calibration (Neutron detectors excluded) 3.3.1.11 3.3.1.10 Channel Calibration (Include resistance temperature detector) 3.3.1.12 Perform COT 3.3.1.13 3.3.1.13 Perform TADOT 3.3.1.14 3.3.1.11 Perform TADOT (verification of setpoint is not required) 3.3.1.15 Verify RTS time response 3.3.1.16 ESFAS Instrumentation 3.3.2 3.3.2 Channel Check 3.3.2.1 3.3.2.1 Perform Actuation Logic Test 3.3.2.2 3.3.2.7 Perform Actuation Logic Test (continuity may be excluded) 3.3.2.3 Perform Master Relay Test 3.3.2.4 3.3.2.7 Perform COT 3.3.2.5 3.3.2.2 Perform Slave Relay Test 3.3.2.6 3.3.2.7 Perform TADOT 3.3.2.7 3.3.2.3 Perform TADOT (setpoint verification not required for manual 3.3.2.8 3.3.2.4 functions)
Channel Calibration 3.3.2.9 3.3.2.5 ESFAS Time Response 3.3.2.10 3.3.2.7 Verify pressurizer pressure low and steam line pressure low NOT 3.3.2.6 bypassed PAM Instrumentation 3.3.3 3.3.3 Channel Check 3.3.3.1 3.3.3.1 Channel Calibration 3.3.3.2 3.3.3.2 Remote Shutdown System 3.3.4 Channel Check 3.3.4.1 Verify control circuit and transfer switch functional 3.3.4.2 Channel Calibration 3.3.4.3 Perform TADOT (reactor trip breakers) 3.3.4.4 LOP DG Start Instrumentation 3.3.5 3.3.4 Channel Check 3.3.5.1 Perform TADOT 3.3.5.2 3.3.4.1 Channel Calibration 3.3.5.3 3.3.4.2 Containment Purge and Exhaust Isolation Instrumentation 3.3.6 Channel Check 3.3.6.1 Perform Actuation Logic Test 3.3.6.2
LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-244 Page 3 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 GINNA Perform Master Relay Test 3.3.6.3 Perform Actuation Logic Test (applicable to ESFAS 3.3.6.4 instrumentation)
Perform Master Relay Test 3.3.6.5 Perform COT 3.3.6.6 Perform Slave Relay Test 3.3.6.7 Perform TADOT (validation of setpoint not required) 3.3.6.8 Channel Calibration 3.3.6.9 Containment Ventilation Isolation Instrumentation 3.3.5 Channel Check 3.3.5.1 Perform COT 3.3.5.2 Perform Actuation Logic Test 3.3.5.3 Channel Calibration 3.3.5.4 CREFS Actuation Instrumentation 3.3.7 Channel Check 3.3.7.1 Perform COT 3.3.7.2 Perform Actuation Logic Test 3.3.7.3 Perform Master Relay Test 3.3.7.4 Perform Actuation Logic Test (applicable to ESFAS 3.3.7.5 instrumentation)
Perform Master Relay Test (applicable to ESFAS instrumentation) 3.3.7.6 Perform Slave Relay Test 3.3.7.7 Perform TADOT (validation of setpoint not required) 3.3.7.8 Channel Calibration 3.3.7.9 CREATS Actuation Instrumentation 3.3.6 Channel Check 3.3.6.1 Perform COT 3.3.6.2 Perform TADOT 3.3.6.3 Channel Calibration 3.3.6.4 Perform Actuation Logic Test 3.3.6.5 FBACS Actuation Instrumentation 3.3.8 Channel Check 3.3.8.1 Perform COT 3.3.8.2 Perform Actuation Logic Test 3.3.8.3 Perform TADOT (validation of setpoint not required) 3.3.8.4 Channel Calibration 3.3.8.5 BDPS 3.3.9 Channel Check 3.3.9.1 Perform COT 3.3.9.2 Channel Calibration 3.3.9.3 RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4.1 Verify pressurizer pressure per COLR 3.4.1.1 3.4.1.1 Verify RCS average temperature per COLR 3.4.1.2 3.4.1.2 Verify RCS total flow rate per COLR 3.4.1.3 3.4.1.3 Verify by heat balance that RCS total flow rate per COLR 3.4.1.4 RCS Minimum Temperature for Criticality 3.4.2 3.4.2
LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-244 Page 4 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 GINNA Verify RCS average temperature in each loop Ž:[540] 3.4.2.1 Verify RCS average temperature in each loop ->[540]Only required ------------- 3.4.2.2 if Tave alarm is inoperable and RCS loop Tave <[547]
RCS P/T Limit 3.4.3 3.4.3 Verify RCS pressure, temperature, heat up and cooldown rates per 3.4.3.1 3.4.3.1 PTLR RCS Loops - MODES 1 and 2 3.4.4 Verify each RCS loop is in operation 3.4.4.1 RCS Loops - MODES 1 >8.5% RTP 3.4.4 Verify each RCS loop is in operation 3.4.4.1 RCS Loops - MODES 1 < 8.5% RTP, 2 and 3 3.4.5 Verify required RCS loop is in operation 3.4.5.1 Verify steam generator secondary side water level 3.4.5.2 Verify correct breaker alignment and power to required RCP pump -------------- 3.4.5.3 RCS Loops - MODE 3 3.4.5 Verify required RCS loops are in operation 3.4.5.1 Verify steam generator water level 3.4.5.2 Verify correct breaker alignment and power to pumps 3.4.5.3 RCS Loops - MODE 4 3.4.6 3.4.6 Verify required RCS or RHR loop is in operation 3.4.6.1 3.4.6.1 Verify steam generator water level 3.4.6.2 3.4.6.2 Verify correct breaker alignment and power to pump 3.4.6.3 3.4.6.3 RCS Loops - MODE 5, Loops Filled 3.4.7 3.4.7 Verify required RHR loop is in operation 3.4.7.1 3.4.7.1 Verify steam generator secondary side water level 3.4.7.2 3.4.7.2 Verify correct breaker alignment and power to required RHR pump 3.4.7.3 3.4.7.3 RCS Loops - MODE 5, Loops Not Filled 3.4.8 3.4.8 Verify required RHR loop is in operation 3.4.8.1 3.4.8.1 Verify correct breaker alignment and power to RHR pump 3.4.8.2 3.4.8.2 Pressurizer 3.4.9 3.4.9 Verify pressurizer water level 3.4.9.1 3.4.9.1 Verify capacity of each group of pressurizer heaters 3.4.9.2 3.4.9.2 Verify required pressurizer heaters capable being powered by 3.4.9.3 emergency power Pressurizer PORVs 3.4.11 3.4.11 Cycle each block valve 3.4.11.1 3.4.11.1 Cycle each PORV 3.4.11.2 3.4.11.2 Cycle each solenoid air control valve 3.4.11.3 Verify PORVs and block valves capable being powered by 3.4.11.4 emergency power LTOP System 3.4.12 3.4.12 Verify no SI pump capable of injecting 3.4.12.1 3.4.12.1 3.4.12.2 Verify maximum of [one] [HPI] pump capable of injecting 3.4.12.1 3.4.12.2 3.4.12.2 Verify maximum of charging pump capable of injecting 3.4.12.2 3.4.12.2 Verify each accumulator is isolated 3.4.12.3 3.4.12.3
LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-244 Page 5 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 GINNA Verify RHR suction valve is open 3.4.12.4 Verify RCS vent path 3.4.12.5 3.4.12.4 Verify (required) PORV block valve open 3.4.12.6 3.4.12.5 Verify RHR suction valve is locked open and power removed 3.4.12.7 Perform COT (excluding actuation) 3.4.12.8 3.4.12.6 Verify power removed from each ECCS accumulator MOV 3.4.12.7 Channel Calibration 3.4.12.9 3.4.12.8 RCS Operational Leakage 3.4.13 3.4.13 Verify RCS operational leakage within limits 3.4.13.1 3.4.13.1 Verify primary to secondary leakage within limits 3.4.13.2 3.4.13.2 RCS PIV Leakage 3.4.14 3.4.14 Verify leakage from each RCS PIV 3.4.14.1 3.4.14.1 3.4.14.2 Verify RHR interlock functionality (opening) 3.4.14.2 Verify RHR interlock functionality (closure) 3.4.14.3 RCS Leakage Detection Instrumentation 3.4.15 3.4.15 Channel Check 3.4.15.1 3.4.15.1 Perform COT (atmosphere radioactivity monitor) 3.4.15.2 3.4.15.2 Channel Calibration (containment sump monitor) 3.4.15.3 3.4.15.3 Channel Calibration (containment atmosphere monitor) 3.4.15.4 3.4.15.4 Channel Calibration (containment air cooler flow rate monitor) 3.4.15.5 RCS Specific Activity 3.4.16 3.4.16 Verify gross specific activity 3.4.16.1 3.4.16.1 Verify Dose Equivalent 1-131 3.4.16.2 3.4.16.2 Determine E 3.4.16.3 3.4.16.3 RCS Loop Isolation Valves 3.4.17 Verify loop isolation valve is open and power removed 3.4.17.1 RCS Loops - Test Exceptions 3.4.19 Verify thermal power is <P-7 3.4.19.1 Accumulators 3.5.1 3.5.1 Verify accumulator isolation valves fully open 3.5.1.1 3.5.1.1 Verify borated water volume 3.5.1.2 3.5.1.2 Verify Nitrogen cover pressure 3.5.1.3 3.5.1.3 Verify boron concentration 3.5.1.4 3.5.1.4 Verify power removed from isolation valve 3.5.1.5 3.5.1.5 ECCS - Operating 3.5.2 3.5.2 Verify (listed) valves in proper position with power removed 3.5.2.1 3.5.2.1 Verify valves in flow path in the proper position with power 3.5.2.2 3.5.2.2 removed Verify breakers, for each valve listed in 3.5.2.1, in correct position 3.5.2.3 Verify ECCS piping full of water 3.5.2.3 Verify ECCS automatic valves actuates to its proper position on an 3.5.2.5 3.5.2.5 actual or simulated signal Verify ECCS pumps starts automatically on an actual or simulated 3.5.2.6 3.5.2.6 signal Verify ECCS throttle valves (listed) in the correct position 3.5.2.7 Visual inspection of ECCS train 3.5.2.8 3.5.2.7
LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-244 Page 6 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 GINNA RWST 3.5.4 3.5.4 Verify RWST borated water temperature 3.5.4.1 Verify RWST borated water volume 3.5.4.2 3.5.4.1 Verify RWST boron concentration 3.5.4.3 3.5.4.2 Seal Injection Flow 3.5.5 Verify manual seal injection valves adjusted to proper flow 3.5.5.1 BIT 3.5.6 Verify BIT borated water temperature 3.5.6.1 Verify BIT borated water volume 3.5.6.2 Verify BIT boron concentration 3.5.6.3 Containment Air Locks 3.6.2 3.6.2 Verify one door can be opened at a time 3.6.2.2 3.6.2.2 Containment Isolation Valves / Boundaries 3.6.3 3.6.3 Verify purge valve [42 inch] closed 3.6.3.1 Verify purge valve [8 inch] closed 3.6.3.2 3.6.3.1 Verify manual isolation valves, blind flanges outside containment 3.6.3.3 3.6.3.2 closed Verify isolation times 3.6.3.5 Cycle each testable check valves through one full cycle 3.6.3.6 Leakrate purge valves with resilient seals 3.6.3.7 Verify automatic valves actuate to their correct position on an 3.6.3.8 3.6.3.6 actual or simulated signal Cycle each testable check valves through one full cycle 3.6.3.9 Verify purge valve is blocked to restrict flow 3.6.3.10 Containment Pressure 3.6.4A 3.6.4 Verify containment pressure within limits 3.6.4A.1 3.6.4.1 Containment Temperature 3.6.5A 3.6.5A Verify containment average temperature within limits 3.6.5A.1 3.6.5.1 Containment Spray and Cooling Systems / (CS,CRFC and NaOH 3.6.6A 3.6.6 System)
Verify valves in the flow path in the correct position 3.6.6A.1 3.6.6.2 Operate containment cooling system (->15 minutes) 3.6.6A.2 3.6.6.4 Verify containment cooling train water flow rate(->700 gpm) 3.6.6A.3 3.6.6.5 Verify containment spray valves actuate to their correct position on 3.6.6A.5 3.6.6.10 an actual or simulated signal Verify containment spray pumps starts on an actual or simulated 3.6.6A.6 3.6.6.11 signal Verify containment cooling train starts on an actual or simulated 3.6.6A.7 3.6.6.12 signal Spray Additive System 3.6.7 Verify valves in the flow path in the correct position 3.6.7.1 3.6.6.3 Verify spray tank solution volume 3.6.7.2 3.6.6.7 Verify NaOH concentration 3.6.7.3 3.6.6.8 Verify containment spray additive tank valves actuate to their 3.6.7.4 3.6.6.13 correct position on an actual or simulated signal Verify spray additive flow from each solution flow path 3.6.7.5 3.6.6.14 MSIV / (MSIV and Non-Return Check Valves) 3.7.2 3.7.2
LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-244 Page 7 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 GINNA Verify MSIV actuate to its isolation position on an actual or 3.7.2.2 3.7.2.3 simulated signal MFIVs and MFRVs, and Associated Bypass Valves 3.7.3 3.7.3 Verify the isolation time of each MFIV, MFRV and associated 3.7.3.2 bypass valve Atmosphere Dump (Relief) Valves (ADVs)(ARVs) 3.7.4 3.7.4 Verify once complete cycle of each ADV(ARV) 3.7.4.1 3.7.4.1 Verify once complete cycle of each ADV (ARV) block valve 3.7.4.2 3.7.4.2 AFW System 3.7.5 3.7.5 Verify valves in the water and steam flow path in their correct 3.7.5.1 3.7.5.1 position Verify each AFW automatic valve actuates to the isolation position 3.7.5.3 3.7.5.5 on an actual or simulated signal Verify each AFW pump starts automatically on an actual or 3.7.5.4 3.7.5.6 simulated signal Verify each SAFW train can be operated from control room 3.7.5.7 Condensate Storage Tank (CST) 3.7.6 3.7.6 Verify CST level 3.7.6.1 3.7.6.1 Component Cooling Water System (CCW) 3.7.7 3.7.7 Verify each CCW valve is in the correct position 3.7.7.1 3.7.7.1 Verify each CCW valve in the flow path actuates to the correct 3.7.7.2 position on an actual or simulated signal Verify each CCW pump starts automatically on an actual or 3.7.7.3 simulated signal Service Water System (SWS) 3.7.8 3.7.8 Verify screen house bay water level and temperatures 3.7.8.1 Verify each SWS valve is in the correct position 3.7.8.1 3.7.8.2 Verify each SWS valve in the flow path actuates to the correct 3.7.8.2 3.7.8.4 position on an actual or simulated signal Verify each SWS pump starts automatically on an actual or 3.7.8.3 3.7.8.5 simulated signal Verify SW loop header cross-tie valves in correct position 3.7.8.3 Ultimate Heat Sink (UHS) 3.7.9 Verify water level in the UHS 3.7.9.1 Verify average water temperature in the UHS 3.7.9.2 Operate each cooling tower fan (>15 minutes) 3.7.9.3 Verify cooling fan starts automatically on an actual or simulated 3.7.9.4 signal Control Room Emergency Filtration System (CREFS) 3.7.10 Operate each CREFS train (>15 minutes) with the heaters on (>15 3.7.10.1 minutes)
Verify each CREF train actuates on an actual or simulated signal 3.7.10.3 Verify each CREF train maintain a positive pressure 3.7.10.4 Control Room Emergency Air Temperature Control System 3.7.11 3.7.9 (CREATCS)
Verify the CREATMS removes the assume heat load 3.7.11.1 Operate CREATS filtration train (>15 minutes) 3.7.9.1
LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-244 Page 8 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 GINNA Verify each CREATS train actuates on an actual or simulated signal ------------- 3.7.9.3 Auxiliary Building Ventilation System 3.7.10 Verify ABVS in operation 3.7.10.1 Verify ABVS maintains negative pressure with respect to Aux. Bldg. ------------- 3.7.10.2 ECCS PREACS 3.7.12 Operate each PREAC for ->10hours with heaters on or for -15 3.7.12.1 minutes for systems without heaters Verify each PREAC train actuates on an actual or simulated signal 3.7.12.3 Verify PREAC can maintained pressure 3.7.12.4 Verify each ECCS PREAC filter bypass damper closed 3.7.12.5 FBACS 3.7.13 Operate each FBACS for -Ž10hours with heaters on or for >15 3.7.13.1 minutes for systems without heaters Verify each FBACS train actuates on an actual or simulated signal 3.7.13.3 Verify FBACS can maintained pressure 3.7.13.4 Verify each FBACS filter bypass damper closed 3.7.13.5 PREACS 3.7.14 Operate each PREAC for ->10 hours with heaters on or for Ž15 3.7.14.1 minutes for systems without heaters Verify each PREAC train actuates on an actual or simulated signal 3.7.14.3 Verify PREAC can maintained pressure 3.7.14.4 Verify each ECCS PREAC filter bypass damper closed 3.7.14.5 Fuel Storage Pool Water Level 3.7.15 3.7.11 Verify fuel storage pool water level (-Ž23 feet) 3.7.15.1 3.7.11.1 Spent Fuel Pool Boron Concentration 3.7.16 3.7.12 Verify spent fuel boron concentration within limits 3.7.16.1 3.7.12.1 Secondary Specific Activity 3.7.18 3.7.14 Verify specific activity of Dose Equivalent 1-131 (50.10) 3.7.18.1 3.7.14.1 AC Sources - Operating 3.8.1 3.8.1 Verify correct breaker alignment 3.8.1.1 3.8.1.1 Verify each diesel starts from standby conditions 3.8.1.2 3.8.1.2 Verify each diesel is synchronized and loaded (for - 60 minutes) 3.8.1.3 3.8.1.3 Verify each day tank contains proper fuel quantity (- 220 gallons) 3.8.1.4 3.8.1.4 Check and remove accumulated water 3.8.1.5 Verify fuel oil transfer operation (from storage tank to day tanks) 3.8.1.6 3.8.1.5 Verify each diesel starts from standby conditions in proper time 3.8.1.7 3.8.1.2 (510 sec)
Verify transfer of AC power sources (Normal to Alternate) 3.8.1.8 3.8.1.6 Load rejection test (largest post-accident load) 3.8.1.9 3.8.1.7 Verify diesel does not trip and voltage is maintained during and 3.8.1.10 3.8.1.7 following the load rejection Verify diesel performs properly on an actual or simulated loss of 3.8.1.11 3.8.1.9 offsite power signal Verify on an actual or simulated ESF actuation each diesel auto 3.8.1.12 3.8.1.9 starts from standby conditions Verify non critical trips are bypassed 3.8.1.13 3.8.1.8 Verify each diesel operates for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.8.1.14
LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-244 Page 9 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 GINNA Verify diesel starts and performs properly within 5 minutes of 3.8.1.15 operating for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> a maximum load Verify diesel synchronizes with offsite power while loaded with 3.8.1.16 emergency loads Verify an actual or simulated ESF signal overrides a test signal 3.8.1.17 Verify interval between each sequenced load block 3.8.1.18 Verify on an actual or simulated loss of offsite power in conjunction 3.8.1.19 with an actual or simulated ESF signal the diesel performs properly Verify when started simultaneously from standby conditions each 3.8.1.20 diesel performs properly Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 3.8.3 Verify each fuel oil storage tank volume (gallons) 3.8.3.1 3.8.3.1 Verify lube oil inventory 3.8.3.2 Verify diesel start receiver pressure 3.8.3.4 Check and remove accumulated water 3.8.3.5 DC Sources - Operating 3.8.4 3.8.4 Verify battery terminal voltage 3.8.4.1 3.8.4.1 Verify battery charger can recharge the battery 3.8.4.2 Verify battery capacity 3.8.4.3 3.8.4.2 3.8.4.3 Battery Parameters 3.8.6 3.8.6 Verify battery float current 3.8.6.1 Verify each pilot cell voltage 3.8.6.2 Verify connected batteries electrolyte level 3.8.6.3 3.8.6.1 Verify average electrolyte temperature (fifth cell of each battery) 3.8.6.5 Verify specific gravity of pilot cell 3.8.6.3 Verify specific gravity of each connected cell 3.8.6.6 Verify each pilot cell temperature 3.8.6.4 3.8.6.4 Verify connected battery cell voltage 3.8.6.5 3.8.6.2 Verify battery capacity 3.8.6.6 3.8.4.3 Inverters - Operating 3.8.7 Verify inverter voltage and alignment to AC buses 3.8.7.1 AC Instrument Bus Sources - MODES 1, 2, 3 and 4 3.8.7 Verify correct static switch alignment 3.8.7.1 Verify correct Class 1 E CVT alignment 3.8.7.2 Inverters - Shutdown 3.8.8 Verify inverter voltage and alignment to AC buses 3.8.8.1 AC Instrument Bus Sources - MODES 5 and 6 3.8.8 Verify correct static switch alignment 3.8.8.1 Verify correct Class 1 E CVT alignmnet 3.8.8.2 Distribution Systems - Operating 3.8.9 Verify correct breaker alignment and voltage of required buses 3.8.9.1 Distribution Systems - MODES 1, 2, 3 and 4 3.8.9 Verify correct breaker alignment and voltage of required buses 3.8.9.1 Distribution Systems - Shutdown 3.8.10 Verify correct breaker alignment and voltage of required buses 3.8.10.1 Distribution Systems - MODES 5 and 6 3.8.10
LAR - Adoption of TSTF-425, Revision 3 Attachment 5 Docket No. 50-244 Page 10 of 10 Technical Specification Section Title/Surveillance Description* TSTF-425 GINNA Verify correct breaker alignment and voltage of required buses 3.8.10.1 Boron Concentration 3.9.1 3.9.1 Verify boron concentration within the COLR 3.9.1.1 3.9.1.1 Unborated Water Source Isolation Valves 3.9.2 Verify each valve that isolates unborated water source is closed 3.9.2.1 Nuclear Instrumentation 3.9.3 3.9.2 Channel Check 3.9.3.1 3.9.2.1 Channel Calibration 3.9.3.2 3.9.2.2 Containment Penetrations 3.9.4 3.9.3 Verify containment penetration in required status 3.9.4.1 3.9.3.1 Verify purge and exhaust valves isolate on an actual or simulated 3.9.4.2 3.9.3.2 signal RHR and Coolant Circulation - High Water Level (>23 Ft) 3.9.5 3.9.4 Verify one RHR loop in operation circulating reactor coolant 3.9.5.1 3.9.4.1 RHR and Coolant Circulation - Low Water Level (< 23 Ft) 3.9.6 3.9.5 Verify one RHR loop in operation 3.9.6.1 3.9.5.1 Verify correct breaker alignment and power to operating RHR 3.9.6.2 3.9.5.2 pump Refueling Cavity Water Level 3.9.7 3.9.6 Verify refueling cavity water level above flange 3.9.7.1 3.9.6.1
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- The Technical Specification Section Title/Surveillance Description portion of this attachment is a summary description of the referenced TSTF-425 (NUREG-1431)/Ginna TS Surveillances which is provided for information purposes only and is not intended to be a verbatim description of the TS Surveillances.
ATTACHMENT 6 License Amendment Request R. E. Ginna Nuclear Power Plant Docket No. 50-244 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
Proposed No Significant Hazards Consideration
LAR - Adoption of TSTF-425, Revision 3 Attachment 6 Docket No. 50-244 Page 1 of 2 PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION Description of Amendment Request: This amendment request involves the adoption of approved changes to the standard technical specifications (STS) for Westinghouse Plants, WOG STS (NUREG-1431), to allow relocation of specific TS surveillance frequencies to a licensee-controlled program. The proposed changes are described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (ADAMS Accession No. ML090850642) related to the Relocation of Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b and are described in the Notice of Availability published in the Federal Register on July 6, 2009 (74 FR 31996).
The proposed changes are consistent with NRC-approved Industry/ TSTF Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5b."
The proposed changes relocate surveillance frequencies to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP). The changes are applicable to licensees using probabilistic risk guidelines contained in NRC-approved NEI 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. 071360456).
Basis for proposed no significant hazards consideration: As required by 10 CFR 50.91 (a),
the Exelon analysis of the issue of no significant hazards consideration is presented below:
- 1. Do the proposed changes involve a significant increase in the probability or consequences of any accident previously evaluated?
Response: No.
The proposed changes relocate the specified frequencies for periodic surveillance requirements to licensee control under a new Surveillance Frequency Control Program.
Surveillance frequencies are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The systems and components required by the technical specifications for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements, and be capable of performing any mitigation function assumed in the accident analysis. As a result, the consequences of any accident previously evaluated are not significantly increased.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Do the proposed changes create the possibility of a new or different kind of accident from any previously evaluated?
Response: No.
No new or different accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the
LAR - Adoption of TSTF-425, Revision 3 Attachment 6 Docket No. 50-244 Page 2 of 2 changes do not impose any new or different requirements. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Do the proposed changes involve a significant reduction in the margin of safety?
Response: No.
The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SSCs), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, Exelon will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1, in accordance with the TS SFCP.
NEI 04-10, Rev. 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
Based upon the above, Exelon concludes that the requested changes do not involve a significant hazards consideration as set forth in 10 CFR 50.92(c), "Issuance of Amendment."
ATTACHMENT 7 License Amendment Request R. E. Ginna Nuclear Power Plant Docket No. 50-244 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)
Proposed Inserts
LAR - Adoption of TSTF-425, Revision 3 Attachment 7 Docket No. 50-244 Page 1 of 1 INSERT 1 In accordance with the Surveillance Frequency Control Program INSERT 2 5.5.17 Surveillance Frequency Control proqram This program provides controls for the Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
- a. The Surveillance Frequency Control Program shall contain a list of Frequencies of the Surveillance Requirements for which the Frequency is controlled by the program.
- b. Changes to the Frequency listed in the Surveillance Frequency Controlled Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequency," Revision 1.
- c. The provision of Surveillance Requirement 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
INSERT 3 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.