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& Infectious Waste Management Division of Waste Management South Carolina Department of Health and Environmental Control 2600 Bull St.Columbia, SC 29201 March 31, 2010 Nuclear Regulatory Commission Page 4 OATH AND AFFIRMATION James R. Morris affirms that he is the person who subscribed his name to-the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge. | & Infectious Waste Management Division of Waste Management South Carolina Department of Health and Environmental Control 2600 Bull St.Columbia, SC 29201 March 31, 2010 Nuclear Regulatory Commission Page 4 OATH AND AFFIRMATION James R. Morris affirms that he is the person who subscribed his name to-the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge. | ||
Jame s, Site Vice President Subscribed and sworn to me: Notary Public Date D te hw~My commission expires: | Jame s, Site Vice President Subscribed and sworn to me: Notary Public Date D te hw~My commission expires: | ||
ATTACHMENT 1 DESCRIPTION AND ASSESSMENT Catawba Nuclear Station Attachment 1 Description and Assessment | ATTACHMENT 1 DESCRIPTION AND ASSESSMENT Catawba Nuclear Station Attachment 1 Description and Assessment 1.0 Description The proposed amendment would modify the Catawba Nuclear Station (Catawba)Technical Specifications (TS) by relocating specific 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." Additionally, the change would add a new program, the Surveillance Frequency Control Program, to TS Section 5.0, "Administrative Controls." The changes are consistent with theNRC approved Industry/TSTF Standard Technical Specification (STS) change TSTF-425, Revision 3 (ADAMS Accession No.ML080280275). | ||
Availability of this TSTF was published in the Federal Register notice on July 6, 2009.2.0 Assessment 2.1 Applicability of Published Safety Evaluation Duke has reviewed the Safety Evaluation Report (SER) dated July 6, 2009. This included a review of the NRC staff's evaluation of TSTF-425, Revision 3, and the requirements specified in NEI 04-10, Rev. 1, (ADAMS Accession No. ML071360456). | |||
The proposed amendment would modify the Catawba Nuclear Station (Catawba)Technical Specifications (TS) by relocating specific 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." Additionally, the change would add a new program, the Surveillance Frequency Control Program, to TS Section 5.0, "Administrative Controls." The changes are consistent with theNRC approved Industry/TSTF Standard Technical Specification (STS) change TSTF-425, Revision 3 (ADAMS Accession No.ML080280275). | |||
Availability of this TSTF was published in the Federal Register notice on July 6, 2009.2.0 Assessment | |||
of Published Safety Evaluation Duke has reviewed the Safety Evaluation Report (SER) dated July 6, 2009. This included a review of the NRC staff's evaluation of TSTF-425, Revision 3, and the requirements specified in NEI 04-10, Rev. 1, (ADAMS Accession No. ML071360456). | |||
Attachment 2 includes Duke Energy's documentation with regards to the PRA technical adequacy, consistent with the requirements of Regulatory Guide 1.200, Rev. 1 (ADAMS Accession No. ML070240001), Section 4.2, and describes any PRA models without NRC endorsed standards, including documentation of the quality characteristics of the models.Duke Energy has concluded that the justifications presented in the TSTF-425 proposal and the safety evaluation prepared by the NRC staff is applicable to Catawba Units 1 and 2, and justify this amendment to incorporate the changes to the Catawba TS.2.2 Optional Changes and Variations The proposed amendment is consistent with the STS changes described in TSTF-425, Revision 3, however, Duke Energy proposes variations or deviations from TSTF-425, as identified below.1. The revised (re-typed) | Attachment 2 includes Duke Energy's documentation with regards to the PRA technical adequacy, consistent with the requirements of Regulatory Guide 1.200, Rev. 1 (ADAMS Accession No. ML070240001), Section 4.2, and describes any PRA models without NRC endorsed standards, including documentation of the quality characteristics of the models.Duke Energy has concluded that the justifications presented in the TSTF-425 proposal and the safety evaluation prepared by the NRC staff is applicable to Catawba Units 1 and 2, and justify this amendment to incorporate the changes to the Catawba TS.2.2 Optional Changes and Variations The proposed amendment is consistent with the STS changes described in TSTF-425, Revision 3, however, Duke Energy proposes variations or deviations from TSTF-425, as identified below.1. The revised (re-typed) | ||
TS pages are not included in this proposed amendment due to the number of TS pages affected, the nature of the proposed changes, and the outstanding amendment requests that Catawba currently has under NRC review. Providing only the mark-ups of the proposed TS.changes satisfies the requirements of 10 CFR 50.90. This is an administrative deviation from TSTF-425 with no exceptions to the NRC staff's model safety evaluation dated July 6, 2009. This administrative deviation is consistent with Exelon's Peach Bottom Atomic Power Station License Amendment application dated August 31, 2009 (NRC Accession No. ML092470153). | TS pages are not included in this proposed amendment due to the number of TS pages affected, the nature of the proposed changes, and the outstanding amendment requests that Catawba currently has under NRC review. Providing only the mark-ups of the proposed TS.changes satisfies the requirements of 10 CFR 50.90. This is an administrative deviation from TSTF-425 with no exceptions to the NRC staff's model safety evaluation dated July 6, 2009. This administrative deviation is consistent with Exelon's Peach Bottom Atomic Power Station License Amendment application dated August 31, 2009 (NRC Accession No. ML092470153). | ||
Page 1 of 3 Catawba Nuclear Station 'Attachment 1 Description and Assessment | Page 1 of 3 Catawba Nuclear Station 'Attachment 1 Description and Assessment | ||
: 2. A note in Technical Specification | : 2. A note in Technical Specification 3.3.1 concerning the one-time extension for SR 3.3.1.5 will be deleted since it has expired and the same page is being revised for this amendment. | ||
the one-time extension for SR 3.3.1.5 will be deleted since it has expired and the same page is being revised for this amendment. | |||
: 3. Attachment 5 provides a cross reference table between the NUREG-1431 surveillances included in TSTF-425 versus the Catawba surveillances included in this amendment request. This cross reference table highlights the following: | : 3. Attachment 5 provides a cross reference table between the NUREG-1431 surveillances included in TSTF-425 versus the Catawba surveillances included in this amendment request. This cross reference table highlights the following: | ||
: a. TSTF-425 surveillances with identical corresponding Catawba Surveillance numbers, b. TSTF-425 surveillances and corresponding Catawba Surveillances but with differing Surveillance numbers, c. TSTF-425 surveillances that are not contained in the Catawba TS and therefore not applicable, and d. Catawba plant specific surveillances that are not contained in TSTF-425 surveillance mark-ups, but are applicable to these amendment requests.Concerning the above, Catawba surveillances with identical corresponding TSTF-425 surveillance numbers (item "a" above) are not deviations from TSTF-425.Catawba surveillance numbers that differ from the corresponding TSTF-425 surveillance numbers (item "b" above) are administrative deviations only from TSTF-425 with no impact on the NRC Staff's model SER.TSTF-425 surveillances that are not contained in the Catawba TS (item "c" above) are not applicable to these amendment requests. | : a. TSTF-425 surveillances with identical corresponding Catawba Surveillance numbers, b. TSTF-425 surveillances and corresponding Catawba Surveillances but with differing Surveillance numbers, c. TSTF-425 surveillances that are not contained in the Catawba TS and therefore not applicable, and d. Catawba plant specific surveillances that are not contained in TSTF-425 surveillance mark-ups, but are applicable to these amendment requests.Concerning the above, Catawba surveillances with identical corresponding TSTF-425 surveillance numbers (item "a" above) are not deviations from TSTF-425.Catawba surveillance numbers that differ from the corresponding TSTF-425 surveillance numbers (item "b" above) are administrative deviations only from TSTF-425 with no impact on the NRC Staff's model SER.TSTF-425 surveillances that are not contained in the Catawba TS (item "c" above) are not applicable to these amendment requests. | ||
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Because of the broad scope of potential Initiative 5b applications, and the fact that the impact of assumptions may differ for each surveillance requirement being evaluated, Duke Energy will address each of the deviations from Capability Category II listed in Table 2-1 for the Catawba PRA respectively for each application of Initiative 5b on an application specific basis. Again, if a requirement is not met a justification of why it is acceptable that the requirement has not been met will be provided. | Because of the broad scope of potential Initiative 5b applications, and the fact that the impact of assumptions may differ for each surveillance requirement being evaluated, Duke Energy will address each of the deviations from Capability Category II listed in Table 2-1 for the Catawba PRA respectively for each application of Initiative 5b on an application specific basis. Again, if a requirement is not met a justification of why it is acceptable that the requirement has not been met will be provided. | ||
These results will be with the documentation package for the specific Initiative 5b application. | These results will be with the documentation package for the specific Initiative 5b application. | ||
2.3.4 Methodology to be Used for Initiative 5b Page 6 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 NEI 04-10 Revision 1 provides the detailed process requirements for controlling surveillance frequencies (SF) of the TS Surveillance Requirements (SRs) that have been relocated from the TSs to the SFCP. The methodology described in NEI 04-10 Revision 1 provides a risk-informed process to support a plant expert panel (called an Integrated Decisionmaking Panel or IDP) assessment of proposed changes to SF, assuring appropriate consideration of risk insights and other deterministic factors, which may impact SF, along with appropriate performance monitoring of changes and documentation requirements. | |||
to be Used for Initiative 5b Page 6 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 NEI 04-10 Revision 1 provides the detailed process requirements for controlling surveillance frequencies (SF) of the TS Surveillance Requirements (SRs) that have been relocated from the TSs to the SFCP. The methodology described in NEI 04-10 Revision 1 provides a risk-informed process to support a plant expert panel (called an Integrated Decisionmaking Panel or IDP) assessment of proposed changes to SF, assuring appropriate consideration of risk insights and other deterministic factors, which may impact SF, along with appropriate performance monitoring of changes and documentation requirements. | |||
The Duke Energy SFCP, including the methodology of assessing SF.changes utilized at Catawba, is consistent with NEI 04-10, Revision 1 and the supporting background document TSTF-425-A, Rev. 3 (Ref. 20).2.3.5 Identification of Key Assumptions Identification of Key Assumptions related to SF (if any) and how they will be addressed is given below.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 a SF change is warranted. | The Duke Energy SFCP, including the methodology of assessing SF.changes utilized at Catawba, is consistent with NEI 04-10, Revision 1 and the supporting background document TSTF-425-A, Rev. 3 (Ref. 20).2.3.5 Identification of Key Assumptions Identification of Key Assumptions related to SF (if any) and how they will be addressed is given below.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 a SF 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 SF change impact. Therefore, the methodology requires the performance of selected sensitivity studies on the standby failure rate of the component(s) of interest for the SF change assessment. | The methodology recognizes that a key area of uncertainty for this application is the standby failure rate utilized in the determination of the SF change impact. Therefore, the methodology requires the performance of selected sensitivity studies on the standby failure rate of the component(s) of interest for the SF change assessment. | ||
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This will be done by performing a review of all outstanding departures from Capability Category II against the specific Initiative 5b application being addressed. | This will be done by performing a review of all outstanding departures from Capability Category II against the specific Initiative 5b application being addressed. | ||
The results of this review will be in the documentation package for the specific Initiative 5b application. | The results of this review will be in the documentation package for the specific Initiative 5b application. | ||
2.4 External Events Considerations This section addresses Condition 2 of the NRC Safety Evaluation for Initiative 5b.Specifically it identifies quality characteristics for PRA models for which NRC-endorsed Standards do not exist, consistent with RG 1.200, Sections 1.2 and 1.3, and justifies the methods to be applied for assessing the risk contribution for those sources of risk not addressed by PRA models.NRC endorsed standards currently exist for external hazards including seismic and fire PRAs. Revision 2 of Regulatory Guide (RG) 1.200 (Ref. 21), references the ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 (Ref. 7) for internal and external hazards. An NRC endorsed standard does not currently exist for shutdown PRAs. NEI 04-10 Revision 1 references RG 1.200 Revision 1 and ASME PRA Standard RA-Sb-2005b as the governing documents for Initiative 5b.The NEI 04-10 Revision 1 methodology allows for SF change evaluations to be performed in the absence of quantifiable PRA models for all external hazards. For those cases where the SF 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. In general, it is not expected that seismic, fire, or other external hazards will play a significant role in the impact of a given surveillance frequency change.This section discusses the Catawba overall external hazards analysis methodology, the Catawba specific seismic and fire PRAs, and describes the methodology to be used to address shutdown risk impacts for Initiative 5b consistent with the requirements of the NEI 04-10 Revision 1 methodology. | |||
2.4.1 Overall External Hazards Analysis Methodology The general approach used to develop the external event PRA at Catawba is as follows: 1) Identify all natural and man-made credible external events that may affect the site using many reference sources.Page 8 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 2) A screening analysis was conducted using defined bounding criteria in order to select those events that may require further review.3) A scoping analysis was performed on the remaining non-screened events to determine those that warranted a detailed site and plant-specific analysis.This approach is consistent withthat previously submitted to the NRC in Section 2.3 of Reference 14 and Volume 1, Section 3.0 of Reference | |||
Events Considerations This section addresses Condition 2 of the NRC Safety Evaluation for Initiative 5b.Specifically it identifies quality characteristics for PRA models for which NRC-endorsed Standards do not exist, consistent with RG 1.200, Sections 1.2 and 1.3, and justifies the methods to be applied for assessing the risk contribution for those sources of risk not addressed by PRA models.NRC endorsed standards currently exist for external hazards including seismic and fire PRAs. Revision 2 of Regulatory Guide (RG) 1.200 (Ref. 21), references the ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 (Ref. 7) for internal and external hazards. An NRC endorsed standard does not currently exist for shutdown PRAs. NEI 04-10 Revision 1 references RG 1.200 Revision 1 and ASME PRA Standard RA-Sb-2005b as the governing documents for Initiative 5b.The NEI 04-10 Revision 1 methodology allows for SF change evaluations to be performed in the absence of quantifiable PRA models for all external hazards. For those cases where the SF 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. In general, it is not expected that seismic, fire, or other external hazards will play a significant role in the impact of a given surveillance frequency change.This section discusses the Catawba overall external hazards analysis methodology, the Catawba specific seismic and fire PRAs, and describes the methodology to be used to address shutdown risk impacts for Initiative 5b consistent with the requirements of the NEI 04-10 Revision 1 methodology. | |||
External Hazards Analysis Methodology The general approach used to develop the external event PRA at Catawba is as follows: 1) Identify all natural and man-made credible external events that may affect the site using many reference sources.Page 8 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 2) A screening analysis was conducted using defined bounding criteria in order to select those events that may require further review.3) A scoping analysis was performed on the remaining non-screened events to determine those that warranted a detailed site and plant-specific analysis.This approach is consistent withthat previously submitted to the NRC in Section 2.3 of Reference 14 and Volume 1, Section 3.0 of Reference | |||
: 11. These references provide a greater level of detail of the approach if needed.2.4.2 Catawba Seismic PRA Model As noted in the IPEEE submittal (Ref. 14), Catawba Unit 2 was selected for a trial assessment of the EPRI developed Seismic Margin Methodology, the methodology for assessing the ability of nuclear plants to withstand earthquakes beyond design basis.The assessment established that Catawba would survive earthquake loads up to approximately twice its design basis. This work is documented in EPRI NP-6359 (Ref.22).The current Catawba seismic PRA model of record was last updated as part of Revision 3 of the, PRA model (Ref. 23). However, the current methodology used is the same as that described in detail in the IPE submittal (Ref. 11) and Section 3 of the IPEEE submittal (Ref. 14) both of which have already been reviewed by the NRC. The reader is referred to those references for additional details of the seismic analysis.The plant-specific seismic PRA analysis consists of four steps each of which are described below: 1) The Catawba site was evaluated to obtain the seismic hazard in terms of the frequency of occurrence of ground motions of various magnitudes. | : 11. These references provide a greater level of detail of the approach if needed.2.4.2 Catawba Seismic PRA Model As noted in the IPEEE submittal (Ref. 14), Catawba Unit 2 was selected for a trial assessment of the EPRI developed Seismic Margin Methodology, the methodology for assessing the ability of nuclear plants to withstand earthquakes beyond design basis.The assessment established that Catawba would survive earthquake loads up to approximately twice its design basis. This work is documented in EPRI NP-6359 (Ref.22).The current Catawba seismic PRA model of record was last updated as part of Revision 3 of the, PRA model (Ref. 23). However, the current methodology used is the same as that described in detail in the IPE submittal (Ref. 11) and Section 3 of the IPEEE submittal (Ref. 14) both of which have already been reviewed by the NRC. The reader is referred to those references for additional details of the seismic analysis.The plant-specific seismic PRA analysis consists of four steps each of which are described below: 1) The Catawba site was evaluated to obtain the seismic hazard in terms of the frequency of occurrence of ground motions of various magnitudes. | ||
The site-specific hazard analysis (Ref. 24) was performed using the Seismicity Owners Group (SOG) methodology developed by EPRI for seismic hazard analysis of nuclear power plant sites in the Central and Eastern United States (CEUS).Uncertainties were addressed in the hazard analysis.2) From the site-specific seismic hazard curve, the capacities of important plant structures and equipment to withstand seismic events were evaluated to determine conditional probabilities of failure as a function of ground acceleration for significant contributors (i.e., SSCs). These are commonly referred to as 'fragilities' or the site-specific fragility curves. Plant walkdowns were conducted, the most recent ones consistent with the guidelines of EPRI NP-6041 (Ref. 25).3) An event tree was developed along with supporting top logic and system fault trees to reflect plant response to seismic events. These modified logic models were then solved to obtain Boolean expressions for the seismic event sequences of interest.4) The Boolean expressions were quantified by convolving the probabilistic site Page 9 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 seismicity and the fragilities for the plant structures and equipment obtained in steps 1 and 2. The resulting sequence frequencies are then integrated into the overall Catawba PRA risk results, resulting in final quantitative results.The major changes to the current seismic analysis that have been made since the IPEEE submittal are as follows: 1. Comprehensive review and revision of the seismic analysis documentation write-up.2. Added component/structure fragility information to support values used in analysis.3. Updated model with new Human Reliability Analysis (HRA) data.4. Updated model with new common cause data.5. Changes made to the fault tree are listed below.* Made a new top gate for the model to address containment safeguards responses. | The site-specific hazard analysis (Ref. 24) was performed using the Seismicity Owners Group (SOG) methodology developed by EPRI for seismic hazard analysis of nuclear power plant sites in the Central and Eastern United States (CEUS).Uncertainties were addressed in the hazard analysis.2) From the site-specific seismic hazard curve, the capacities of important plant structures and equipment to withstand seismic events were evaluated to determine conditional probabilities of failure as a function of ground acceleration for significant contributors (i.e., SSCs). These are commonly referred to as 'fragilities' or the site-specific fragility curves. Plant walkdowns were conducted, the most recent ones consistent with the guidelines of EPRI NP-6041 (Ref. 25).3) An event tree was developed along with supporting top logic and system fault trees to reflect plant response to seismic events. These modified logic models were then solved to obtain Boolean expressions for the seismic event sequences of interest.4) The Boolean expressions were quantified by convolving the probabilistic site Page 9 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 seismicity and the fragilities for the plant structures and equipment obtained in steps 1 and 2. The resulting sequence frequencies are then integrated into the overall Catawba PRA risk results, resulting in final quantitative results.The major changes to the current seismic analysis that have been made since the IPEEE submittal are as follows: 1. Comprehensive review and revision of the seismic analysis documentation write-up.2. Added component/structure fragility information to support values used in analysis.3. Updated model with new Human Reliability Analysis (HRA) data.4. Updated model with new common cause data.5. Changes made to the fault tree are listed below.* Made a new top gate for the model to address containment safeguards responses. | ||
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Duke Energy is planning to perform a self-assessment against the supporting requirements for fire events of ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 for the Catawba fire PRA in 2010. The method as described in Section 2.3.3 of this attachment will be used to justify any departures from the ASME Standard Capability Category II requirements for each application of Initiative 5b. However, in accordance with the discussion in this section above, Duke Energy considers the current fire model of record as meeting the required quality characteristics of RG 1.200 Sections 1.2 and 1.3 and is therefore sufficient for use as is in the application of Initiative 5b SF changes.2.4.3.1 Catawba Future State Fire PRA Model Initiative In February 2005, Duke Energy notified the NRC (Ref. 27) of its intent to adopt National Fire Protection Association (NFPA) Standard 805, "NFPA 805, Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generation Plants," 2001 edition, pursuant to Section 50.48(c) of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50.48(c)), at all of its nuclear stations.In a letter dated June 8, 2005, the NRC accepted Duke Energy's intent to adopt 10 CFR 50.48(c) (NFPA 805 Rule) for all three sites with Oconee Nuclear Station beginning the transition as a pilot plant on June 1, 2005 (Ref. 28). Duke Energy was requested to inform the NRC when the transition would begin at Catawba.Subsequently, Duke Energy informed the NRC in 2007 (Ref. 29) that the transition to NFPA 805 at Catawba Nuclear Station had begun. The NRC response on January 4, 2008 (Ref. 30) acknowledged the transition to the performance-based standard for fire protection had begun at Catawba Units 1 and 2.The Catawba Fire PRA model being developed uses guidance contained in NUREG/CR-6850/EPRI TR-1011989, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities (Ref. 31). This is the same methodology and approach as that being used for the Oconee pilot. The Catawba Fire PRA model is to receive an industry peer review-against the requirements of Part 4 of ASME/ANS RA-Sa-2009, Addendum to RA-S-2008 (Ref. 7) in April 2010. When the'peer review report is received the departures from Capability Category II requirements and other findings will be addressed. | Duke Energy is planning to perform a self-assessment against the supporting requirements for fire events of ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 for the Catawba fire PRA in 2010. The method as described in Section 2.3.3 of this attachment will be used to justify any departures from the ASME Standard Capability Category II requirements for each application of Initiative 5b. However, in accordance with the discussion in this section above, Duke Energy considers the current fire model of record as meeting the required quality characteristics of RG 1.200 Sections 1.2 and 1.3 and is therefore sufficient for use as is in the application of Initiative 5b SF changes.2.4.3.1 Catawba Future State Fire PRA Model Initiative In February 2005, Duke Energy notified the NRC (Ref. 27) of its intent to adopt National Fire Protection Association (NFPA) Standard 805, "NFPA 805, Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generation Plants," 2001 edition, pursuant to Section 50.48(c) of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50.48(c)), at all of its nuclear stations.In a letter dated June 8, 2005, the NRC accepted Duke Energy's intent to adopt 10 CFR 50.48(c) (NFPA 805 Rule) for all three sites with Oconee Nuclear Station beginning the transition as a pilot plant on June 1, 2005 (Ref. 28). Duke Energy was requested to inform the NRC when the transition would begin at Catawba.Subsequently, Duke Energy informed the NRC in 2007 (Ref. 29) that the transition to NFPA 805 at Catawba Nuclear Station had begun. The NRC response on January 4, 2008 (Ref. 30) acknowledged the transition to the performance-based standard for fire protection had begun at Catawba Units 1 and 2.The Catawba Fire PRA model being developed uses guidance contained in NUREG/CR-6850/EPRI TR-1011989, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities (Ref. 31). This is the same methodology and approach as that being used for the Oconee pilot. The Catawba Fire PRA model is to receive an industry peer review-against the requirements of Part 4 of ASME/ANS RA-Sa-2009, Addendum to RA-S-2008 (Ref. 7) in April 2010. When the'peer review report is received the departures from Capability Category II requirements and other findings will be addressed. | ||
In September 2010, Duke Energy is planning to submit a License Amendment Request (LAR) to the NRC to adopt the new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a), 10 CFR 50.48(c), and the guidance in Regulatory Guide (RG) 1.205. A discussion of the peer review open items and their disposition is expected to be part of that submittal. | In September 2010, Duke Energy is planning to submit a License Amendment Request (LAR) to the NRC to adopt the new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a), 10 CFR 50.48(c), and the guidance in Regulatory Guide (RG) 1.205. A discussion of the peer review open items and their disposition is expected to be part of that submittal. | ||
2.4.4 Catawba Shutdown Risk Impact Analysis Since no approved quantitative shutdown risk PRA model for shutdown events currently exists at Duke Energy, Catawba will either 1) utilize the plant shutdown safety assessment tool developed to support implementation of NUMARC 91-06 (Ref. 32) as Page 12 of 27 Catawba Nuclear Station Attachment 2-Adoption of TSTF-425, Revision 3 described in Duke Energy Nuclear Station Directive (NSD) 403 (Ref. 33) or 2) perform an alternate qualitative risk evaluation process to assess the proposed surveillance frequency change that utilizes Initiative 5b. These are acceptable options to not having a quantitative shutdown PRA model in accordance with Section 4 Step 10 (and other places) of NEI 04-10 Revision 1. In either case, the guidance of NEI 04-10 Revision 1 will be followed.2.5 Summary In Section 2.3 of this document the Catawba PRA technical adequacy was evaluated in accordance with the requirements of RG 1.200, Section 4.2. Section 2.4 of this document submitted quality characteristics of the seismic and fire PRA models in accordance with the requirements of RG 1.200, Sections 1.2 and 1.3. A discussion of the qualitative method to address shutdown risk was also discussed in Section 2.4.Because of the broad scope of potential Initiative 5b applications and the fact that the risk assessment details will differ from application to application, for each individual SF interval request, a review of the unincorporated changes to the plant and remaining gaps to specific requirements in the PRA standard will be made to determine which, if any, would merit additional application-specific sensitivity studies in the final analysis.The results of the discussions above provide a basis for concluding that the current Catawba Units 1 and 2 PRA model is sufficiently robust and suitable for use in risk-informed processes such as that proposed for the implementation of a Surveillance Frequency Control Program.2.6 References | |||
Shutdown Risk Impact Analysis Since no approved quantitative shutdown risk PRA model for shutdown events currently exists at Duke Energy, Catawba will either 1) utilize the plant shutdown safety assessment tool developed to support implementation of NUMARC 91-06 (Ref. 32) as Page 12 of 27 Catawba Nuclear Station Attachment 2-Adoption of TSTF-425, Revision 3 described in Duke Energy Nuclear Station Directive (NSD) 403 (Ref. 33) or 2) perform an alternate qualitative risk evaluation process to assess the proposed surveillance frequency change that utilizes Initiative 5b. These are acceptable options to not having a quantitative shutdown PRA model in accordance with Section 4 Step 10 (and other places) of NEI 04-10 Revision 1. In either case, the guidance of NEI 04-10 Revision 1 will be followed.2.5 Summary In Section 2.3 of this document the Catawba PRA technical adequacy was evaluated in accordance with the requirements of RG 1.200, Section 4.2. Section 2.4 of this document submitted quality characteristics of the seismic and fire PRA models in accordance with the requirements of RG 1.200, Sections 1.2 and 1.3. A discussion of the qualitative method to address shutdown risk was also discussed in Section 2.4.Because of the broad scope of potential Initiative 5b applications and the fact that the risk assessment details will differ from application to application, for each individual SF interval request, a review of the unincorporated changes to the plant and remaining gaps to specific requirements in the PRA standard will be made to determine which, if any, would merit additional application-specific sensitivity studies in the final analysis.The results of the discussions above provide a basis for concluding that the current Catawba Units 1 and 2 PRA model is sufficiently robust and suitable for use in risk-informed processes such as that proposed for the implementation of a Surveillance Frequency Control Program.2.6 References | |||
: 1. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities", Revision 1, US Nuclear Regulatory Commission, January 2007.2. ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications", with Addenda ASME RA-Sa-2003 and ASME RA-Sb-2005, December 2005.3. NEI 00-02, "Probabilistic Risk Assessment Peer Review Process Guidance," Revision A3, Nuclear Energy Institute, March 20, 2000.4. Letter, USNRC to Nuclear Energy Institute, "Final Safety Evaluation for Nuclear Energy Institute (NEI) Industry Guidance Document NEI 04-10, Revision 0, "Risk-Informed Technical Specifications Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies"", September 28, 2006.5. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," April 2007.6. DPC-1535.00-00-0013 (Cross references: | : 1. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities", Revision 1, US Nuclear Regulatory Commission, January 2007.2. ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications", with Addenda ASME RA-Sa-2003 and ASME RA-Sb-2005, December 2005.3. NEI 00-02, "Probabilistic Risk Assessment Peer Review Process Guidance," Revision A3, Nuclear Energy Institute, March 20, 2000.4. Letter, USNRC to Nuclear Energy Institute, "Final Safety Evaluation for Nuclear Energy Institute (NEI) Industry Guidance Document NEI 04-10, Revision 0, "Risk-Informed Technical Specifications Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies"", September 28, 2006.5. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," April 2007.6. DPC-1535.00-00-0013 (Cross references: | ||
CNC-1 535.00-00-0094, MCC-1535.00-00-0089, OSC-9380), "PRA Quality Self-Assessment, Catawba Units 1 & 2, McGuire Units I & 2, Oconee Units 1, 2 & 3", Revision 2, November 2009.7. ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.Page 13 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 8. "Catawba Nuclear Station, Unit 1 Probabilistic Risk Assessment, "Volume 1, Preface, Duke Power Company, August 1987.9. "Catawba Nuclear Station Unit 1 Probabilistic RiskAssessment," Volumes 1-3, Duke Power Company, August 18, 1987.10. NRC Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities", US Nuclear Regulatory Commission, November 23, 1988.11. Letter Duke Power Company to Document Control Desk (USNRC), Catawba Units 1 and 2, "Individual Plant Examination (IPE) Submittal in Response to Generic Letter 88-20," September 10, 1992.12. Letter USNRC to Duke Power Company, "Safety Evaluation of Catawba Nuclear Station, Units 1 and 2 Individual Plant Examination (IPE) Submittal," June 7, 1994.13. 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..14. Letter Duke Power Company to Document Control Desk (USNRC), Catawba Units 1 and 2, "Individual Plant Examination of External Events (IPEEE) Submittal," June 21, 1994.15. Letter Duke Power Company to Document Control Desk (USNRC), Catawba Nuclear Station, "Request for Additional Information-Individual Plant Examinations for External Events; Response," November 17, 1995.16. Letter Duke Power Company to Document Control Desk (USNRC), "Supplemental IPEEE Report," Duke Power Company, McGuire Nuclear Station, Catawba Nuclear Station, July 30, 1996.17: Letter USNRC to Duke Power Company, "Catawba Nuclear Station -Review of Individual Plant Examination of External Events (IPEEE)," April 12, 1999.18. Letter Duke Energy Corporation to Document Control Desk (USNRC), Catawba Units 1 and 2, "Probabilistic Risk Assessment (PRA), Revision 2 Summary Report, January 1998.19. "Catawba Nuclear Station Probabilistic Safety Assessment Peer Review Report", Westinghouse Electric Co. for the Westinghouse Owners Group, December 2002.20. Technical Specification Task Force Traveler number TSTF-425, Revision 3,"Relocate Surveillance Frequencies to Licensee Control -RITSTF Initiative 5," July 2009.21. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities", Revision 2, US Nuclear Regulatory Commission, March 2009.22. EPRI NP-6359, "Seismic Margin Assessment of the Catawba Nuclear Station," April 1989.23. CNC-1535.00-00-0059, Catawba Nuclear Station, External Events -Seismic Analysis, December 2003.24. EPRI NP-4726-A, "Seismic Hazard Methodology for the Central and Eastern United States," July 1986.25. EPRI NP-6041, Revision 1, "A Methodology of Assessment of Nuclear Power Plant Seismic Margin, "August 1991.26. CNC-1535.00-00.0057, Catawba Nuclear Station, Fire Analysis Notebook, September 1997.27. Letter Duke Energy Corporation to Document Control Desk (USNRC), "Letter of Intent to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Page 14 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 Light Water Reactor Generating Plants, 2001 Edition," February 28, 2005 (Adams Accession Number ML050670305). | CNC-1 535.00-00-0094, MCC-1535.00-00-0089, OSC-9380), "PRA Quality Self-Assessment, Catawba Units 1 & 2, McGuire Units I & 2, Oconee Units 1, 2 & 3", Revision 2, November 2009.7. ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.Page 13 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 8. "Catawba Nuclear Station, Unit 1 Probabilistic Risk Assessment, "Volume 1, Preface, Duke Power Company, August 1987.9. "Catawba Nuclear Station Unit 1 Probabilistic RiskAssessment," Volumes 1-3, Duke Power Company, August 18, 1987.10. NRC Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities", US Nuclear Regulatory Commission, November 23, 1988.11. Letter Duke Power Company to Document Control Desk (USNRC), Catawba Units 1 and 2, "Individual Plant Examination (IPE) Submittal in Response to Generic Letter 88-20," September 10, 1992.12. Letter USNRC to Duke Power Company, "Safety Evaluation of Catawba Nuclear Station, Units 1 and 2 Individual Plant Examination (IPE) Submittal," June 7, 1994.13. 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..14. Letter Duke Power Company to Document Control Desk (USNRC), Catawba Units 1 and 2, "Individual Plant Examination of External Events (IPEEE) Submittal," June 21, 1994.15. Letter Duke Power Company to Document Control Desk (USNRC), Catawba Nuclear Station, "Request for Additional Information-Individual Plant Examinations for External Events; Response," November 17, 1995.16. Letter Duke Power Company to Document Control Desk (USNRC), "Supplemental IPEEE Report," Duke Power Company, McGuire Nuclear Station, Catawba Nuclear Station, July 30, 1996.17: Letter USNRC to Duke Power Company, "Catawba Nuclear Station -Review of Individual Plant Examination of External Events (IPEEE)," April 12, 1999.18. Letter Duke Energy Corporation to Document Control Desk (USNRC), Catawba Units 1 and 2, "Probabilistic Risk Assessment (PRA), Revision 2 Summary Report, January 1998.19. "Catawba Nuclear Station Probabilistic Safety Assessment Peer Review Report", Westinghouse Electric Co. for the Westinghouse Owners Group, December 2002.20. Technical Specification Task Force Traveler number TSTF-425, Revision 3,"Relocate Surveillance Frequencies to Licensee Control -RITSTF Initiative 5," July 2009.21. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities", Revision 2, US Nuclear Regulatory Commission, March 2009.22. EPRI NP-6359, "Seismic Margin Assessment of the Catawba Nuclear Station," April 1989.23. CNC-1535.00-00-0059, Catawba Nuclear Station, External Events -Seismic Analysis, December 2003.24. EPRI NP-4726-A, "Seismic Hazard Methodology for the Central and Eastern United States," July 1986.25. EPRI NP-6041, Revision 1, "A Methodology of Assessment of Nuclear Power Plant Seismic Margin, "August 1991.26. CNC-1535.00-00.0057, Catawba Nuclear Station, Fire Analysis Notebook, September 1997.27. Letter Duke Energy Corporation to Document Control Desk (USNRC), "Letter of Intent to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Page 14 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 Light Water Reactor Generating Plants, 2001 Edition," February 28, 2005 (Adams Accession Number ML050670305). | ||
| Line 205: | Line 185: | ||
documentation. | documentation. | ||
1, j Page 27 of 27 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATION CHANGES INSERTS Insert I In accordance with the Surveillance Frequency Control Program REVIEWER'S NOTE: Text deleted and replaced by Insert I Will be relocated to the Surveillance Frequency Control Program (SFCP) document(s) per TSTF-425.Insert 3 5.5.17 Surveillance Freauency Control Program This Program provides controls for Surveillance Frequencies. | 1, j Page 27 of 27 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATION CHANGES INSERTS Insert I In accordance with the Surveillance Frequency Control Program REVIEWER'S NOTE: Text deleted and replaced by Insert I Will be relocated to the Surveillance Frequency Control Program (SFCP) document(s) per TSTF-425.Insert 3 5.5.17 Surveillance Freauency Control Program This Program provides controls for Surveillance Frequencies. | ||
The program shall ensure that the Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operations are met.a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.Note: Insert 2 is included on Attachment | The program shall ensure that the Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operations are met.a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.Note: Insert 2 is included on Attachment | ||
: 4. | : 4. | ||
Definitions 1.1 1.1 Definitions (continued) | Definitions 1.1 1.1 Definitions (continued) | ||
| Line 214: | Line 194: | ||
hannels, or other /1 designate components durin e interval specifie y the Surveill 'ce Frequency, so t~at all systems, subs tems, /chan Is, or other designa/td components are. sted during/n S drveillance Frequend yintervals, where niche total .0Umber of systems, s usystems, channels zr other /designated componefts in the associate unction.THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.THERMAL POWER (continued) | hannels, or other /1 designate components durin e interval specifie y the Surveill 'ce Frequency, so t~at all systems, subs tems, /chan Is, or other designa/td components are. sted during/n S drveillance Frequend yintervals, where niche total .0Umber of systems, s usystems, channels zr other /designated componefts in the associate unction.THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.THERMAL POWER (continued) | ||
Amendment Nos. Catawba Units 1 and 2 1.1-5 SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)LCO 3.1.1 APPLICABILITY: | Amendment Nos. Catawba Units 1 and 2 1.1-5 SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)LCO 3.1.1 APPLICABILITY: | ||
SDM shall be within the limit specified in the COLR.MODE 2 with ke, < 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 limit specified in the COLR. 2 Catawba Units 1 and 2 3.1.1 -1 Amendment Nos. Gýo Core Reactivity | SDM shall be within the limit specified in the COLR.MODE 2 with ke, < 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 limit specified in the COLR. 2 Catawba Units 1 and 2 3.1.1 -1 Amendment Nos. Gýo Core Reactivity 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1------------- | ||
REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1------------- | |||
NOTE -------------------- | NOTE -------------------- | ||
The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.Verify measured core reactivity is within _ 1% Ak/k of predicted values.Once prior to entering MODE 1 after each refueling AND/2?1 Amendment Nos.(7 Catawba Units 1 and 2 3.1.2-2 Rod Group Alignment Limits 3.1.4 ACTIONS (continued) | The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.Verify measured core reactivity is within _ 1% Ak/k of predicted values.Once prior to entering MODE 1 after each refueling AND/2?1 Amendment Nos.(7 Catawba Units 1 and 2 3.1.2-2 Rod Group Alignment Limits 3.1.4 ACTIONS (continued) | ||
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Catawba Units 1 and 2 Amendment 3.1.4-4 Shutdown Bank Insertion Limits 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown bank is within the limits specified in the COLR.Catawba Units 1 and 2 Amendment Nos. 1 3.1.5-2 Control Bank Insertion Limits 3.1.6 SURVEILLANCE REQUIREMENTS (continued) | Catawba Units 1 and 2 Amendment 3.1.4-4 Shutdown Bank Insertion Limits 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown bank is within the limits specified in the COLR.Catawba Units 1 and 2 Amendment Nos. 1 3.1.5-2 Control Bank Insertion Limits 3.1.6 SURVEILLANCE REQUIREMENTS (continued) | ||
SURVEILLANCE FREQUENCY SR 3.1.6.2 Verify each control bank insertion is within the limits specified in the COLR.AND 5 C 17 Once within 4 hours and every 4 hours thereafter when the rod insertion limit monitor is inoperable SR 3.1.6.3 Verify sequence and overlap limits specified in the COLR are met for control banks not fully withdrawn from the core.Eli.Catawba Units 1 and 2 3.1.6-3 Amendment Nos. | SURVEILLANCE FREQUENCY SR 3.1.6.2 Verify each control bank insertion is within the limits specified in the COLR.AND 5 C 17 Once within 4 hours and every 4 hours thereafter when the rod insertion limit monitor is inoperable SR 3.1.6.3 Verify sequence and overlap limits specified in the COLR are met for control banks not fully withdrawn from the core.Eli.Catawba Units 1 and 2 3.1.6-3 Amendment Nos. | ||
PHYSICS TESTS Exceptions | PHYSICS TESTS Exceptions 3.1.8 ACTIONS (continued) | ||
(continued) | |||
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 CHANNEL OPERATIONAL TEST on power Prior to initiation of range and intermediate range channels per SR 3.3.1.7, PHYSICS TESTS SR 3.3.1.8, and Table 3.3.1-1.SR 3.1.8.2 Verify the RCS lowest loop average temperature is [jndtE > 541 'F.SR 3.1.8.3 Verify THERMAL POWER is < 5% RTP. 0_SR 3.1.8.4 Verify SDM is within the limit specified in the COLR. K Catawba Units 1 and 2 3.1.8-2 Amendment FQ(X,Y,Z)3.2.1 SURVEILLANCE REQUIREMENTS | 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 CHANNEL OPERATIONAL TEST on power Prior to initiation of range and intermediate range channels per SR 3.3.1.7, PHYSICS TESTS SR 3.3.1.8, and Table 3.3.1-1.SR 3.1.8.2 Verify the RCS lowest loop average temperature is [jndtE > 541 'F.SR 3.1.8.3 Verify THERMAL POWER is < 5% RTP. 0_SR 3.1.8.4 Verify SDM is within the limit specified in the COLR. K Catawba Units 1 and 2 3.1.8-2 Amendment FQ(X,Y,Z)3.2.1 SURVEILLANCE REQUIREMENTS | ||
---------------------------- | ---------------------------- | ||
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: 1. Adjust NIS channel if absolute difference i~s > 3%.2. Not required to be performed until 24 hours after THERMAL POWER is > 15% RTP.('13 er EF)Compare results of the incore detector measurements to NIS AFD.(continued) | : 1. Adjust NIS channel if absolute difference i~s > 3%.2. Not required to be performed until 24 hours after THERMAL POWER is > 15% RTP.('13 er EF)Compare results of the incore detector measurements to NIS AFD.(continued) | ||
Catawba Units 1. and 2 3.3.1-9 Amendment Nos. , | Catawba Units 1. and 2 3.3.1-9 Amendment Nos. , | ||
RTS Instrumentation | RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued) | ||
REQUIREMENTS (continued) | |||
SURVEILLANCE FREQUENCY SR 3.3.1.4 This Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service.Perform TADOT.SR 3.3.1.5 Perform ACTUATION LOGIC TEST.SR 3.3.1.6-------- ------------------------ | SURVEILLANCE FREQUENCY SR 3.3.1.4 This Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service.Perform TADOT.SR 3.3.1.5 Perform ACTUATION LOGIC TEST.SR 3.3.1.6-------- ------------------------ | ||
..I r~ -----------... | ..I r~ -----------... | ||
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Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours after entry into MODE 3.Perform COT.(continued) | Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours after entry into MODE 3.Perform COT.(continued) | ||
-----( A ' pS T B S " a s it p i o U i* The SR 3.3.1 .- equency of "days on a STAGG -ED TEST BASIS" as itapplies to Unit 2 Train 2A agnTrain 2B reapci trip breaker testin ay be extended on aroaicme basis to March 102009 at 0500 -rs, upon which Uni 2 shall be in Mode 3 wi~treactor trip brears open, r the End of C e 16 Refueling O ge. Upon entry into M6d4 3 with reactor tri br kers open for is refueling outage, his extension shall expird The provisions, R 3.0.2, rnot applica to this extension.,, -- -. -Catawba Units 1 and 2 3.3.1-10 Amendment Nos(ý) | -----( A ' pS T B S " a s it p i o U i* The SR 3.3.1 .- equency of "days on a STAGG -ED TEST BASIS" as itapplies to Unit 2 Train 2A agnTrain 2B reapci trip breaker testin ay be extended on aroaicme basis to March 102009 at 0500 -rs, upon which Uni 2 shall be in Mode 3 wi~treactor trip brears open, r the End of C e 16 Refueling O ge. Upon entry into M6d4 3 with reactor tri br kers open for is refueling outage, his extension shall expird The provisions, R 3.0.2, rnot applica to this extension.,, -- -. -Catawba Units 1 and 2 3.3.1-10 Amendment Nos(ý) | ||
RTS Instrumentation | RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued) | ||
REQUIREMENTS (continued) | |||
SURVEILLANCE FREQUENCY SR 3.3.1.8 This Surveillance shall include verification that interlocks P-6 (for the Intermediate Range channels) and P-10 (for the Power Range channels) are in their required state for existing unit conditions. | SURVEILLANCE FREQUENCY SR 3.3.1.8 This Surveillance shall include verification that interlocks P-6 (for the Intermediate Range channels) and P-10 (for the Power Range channels) are in their required state for existing unit conditions. | ||
Perform COT.i2 1-7--------- | Perform COT.i2 1-7--------- | ||
NOTE -------Only required when not*performed within Prior to reactor startup AND Four hours after reducing power below P-10 for power and intermediate range instrumentation AND Four hours after reducing power below P-6 for source range instrumentation AND (continued) | NOTE -------Only required when not*performed within Prior to reactor startup AND Four hours after reducing power below P-10 for power and intermediate range instrumentation AND Four hours after reducing power below P-6 for source range instrumentation AND (continued) | ||
Catawba Units 1 and 2 3.3.1-11 Amendment Nos 424 RTS Instrumentation | Catawba Units 1 and 2 3.3.1-11 Amendment Nos 424 RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued) | ||
REQUIREMENTS (continued) | |||
SURVEILLANCE FREQUENCY SR 3.3.1.9---------------------------------- | SURVEILLANCE FREQUENCY SR 3.3.1.9---------------------------------- | ||
NOTE ------------------- | NOTE ------------------- | ||
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SR 3.3.1.13 Perform COT.(continued) | SR 3.3.1.13 Perform COT.(continued) | ||
Catawba Units 1 and 2 3.3.1-12 Amendment Nos. . | Catawba Units 1 and 2 3.3.1-12 Amendment Nos. . | ||
RTS Instrumentation | RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued) | ||
REQUIREMENTS (continued) | |||
SURVEILLANCE FREQUENCY SR 3.3.1.14--------------------- | SURVEILLANCE FREQUENCY SR 3.3.1.14--------------------- | ||
NOTE ----------------- | NOTE ----------------- | ||
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......------------- | ......------------- | ||
NOTE -------------------------------- | NOTE -------------------------------- | ||
Neutron detectors are excluded from response time testing.Verify RTS RESPONSE TIME is within limits. 1 __-ns0 SR 3.3.1.17 Verify RTS RESPONSE TIME for RTDs is within limits. 11, _5s, Catawba Units 1 and 2 3.3.1-13 Amendment Nos. (1:7ýj n5, ESFAS Instrumentation | Neutron detectors are excluded from response time testing.Verify RTS RESPONSE TIME is within limits. 1 __-ns0 SR 3.3.1.17 Verify RTS RESPONSE TIME for RTDs is within limits. 11, _5s, Catawba Units 1 and 2 3.3.1-13 Amendment Nos. (1:7ýj n5, ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS | ||
REQUIREMENTS | |||
-------------------------------------------------- | -------------------------------------------------- | ||
------- -NOTE ------------------------------------------------------------ | ------- -NOTE ------------------------------------------------------------ | ||
| Line 357: | Line 310: | ||
Final actuation of pumps or valves not required.Perform TADOT.SR 3.3.2.4 Perform MASTER RELAY TEST.E2~ hour_9-2 days on a STAGGERED TEST BAS 31 days.,!, days on a STAGGERED TEST BASI S da 18 onths for y estinghou AR and Po &Br ieldMDR laytypes SR 3.3.2.5 Perform COT.SR 3.3.2.6 Perform SLAVE RELAY TEST.I SR 3.3.2.7 Perform COT.31 day.jnl,-rrinu~d) | Final actuation of pumps or valves not required.Perform TADOT.SR 3.3.2.4 Perform MASTER RELAY TEST.E2~ hour_9-2 days on a STAGGERED TEST BAS 31 days.,!, days on a STAGGERED TEST BASI S da 18 onths for y estinghou AR and Po &Br ieldMDR laytypes SR 3.3.2.5 Perform COT.SR 3.3.2.6 Perform SLAVE RELAY TEST.I SR 3.3.2.7 Perform COT.31 day.jnl,-rrinu~d) | ||
Catawba Units 1 and 2 3.3.2-10 Amendment No< | Catawba Units 1 and 2 3.3.2-10 Amendment No< | ||
ESFAS Instrumentation | ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS | ||
REQUIREMENTS | |||
-------- --------- | -------- --------- | ||
----------NOTE -------------------------------- | ----------NOTE -------------------------------- | ||
Refer to Table 3.3.2-1 to determine which SRs apply for each ESFAS Function.SR 3.3.2.6 Perform SLAVE RELAY TEST.Catawba Units 1 and 2 3.3.2-10 Amendment NosW ESFAS Instrumentation | Refer to Table 3.3.2-1 to determine which SRs apply for each ESFAS Function.SR 3.3.2.6 Perform SLAVE RELAY TEST.Catawba Units 1 and 2 3.3.2-10 Amendment NosW ESFAS Instrumentation 3.3.2 SURVEILLANCE FREQUENCY SR 3.3.2.8----------------- | ||
FREQUENCY SR 3.3.2.8----------------- | |||
NOTE ................- | NOTE ................- | ||
---------...... | ---------...... | ||
| Line 378: | Line 323: | ||
/ #" SR 3.3.2.10------------------ | / #" SR 3.3.2.10------------------ | ||
NOTE -------------------------------- | NOTE -------------------------------- | ||
Not required to be performed for the turbine driven AFW pump until 24 hours after SG 'pressure is > 600 psig.Verify ESFAS RESPONSE TIMES are within limit.1i8m ths ST ASI SR 3.3.2.11 Perform COT.=8,;wir SR 3.3.2.12 Perform ACTUATION LOGIC TEST.Catawba Units 1 and 2 Amendment Nos. [.3.3.2-1,1 PAM Instrumentation | Not required to be performed for the turbine driven AFW pump until 24 hours after SG 'pressure is > 600 psig.Verify ESFAS RESPONSE TIMES are within limit.1i8m ths ST ASI SR 3.3.2.11 Perform COT.=8,;wir SR 3.3.2.12 Perform ACTUATION LOGIC TEST.Catawba Units 1 and 2 Amendment Nos. [.3.3.2-1,1 PAM Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS N-------------------NC) r --------------------------------------------------------------- | ||
REQUIREMENTS N-------------------NC) r --------------------------------------------------------------- | |||
SR 3.3.3.1 and SR 3.3.3.3 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 instrumentation channel that is normally energized. | SR 3.3.3.1 and SR 3.3.3.3 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 instrumentation channel that is normally energized. | ||
SR 3.3.3.2 Not Used SR 3.3.3.3--------------- | SR 3.3.3.2 Not Used SR 3.3.3.3--------------- | ||
| Line 390: | Line 331: | ||
Catawba Units 1 and 2 3.3.3-3 Amendment Nos. Gý?14 Remote Shutdown System 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform CHANNEL CHECK for each required instrumentation channel that is normally energized. | Catawba Units 1 and 2 3.3.3-3 Amendment Nos. Gý?14 Remote Shutdown System 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform CHANNEL CHECK for each required instrumentation channel that is normally energized. | ||
SR 3.3.4.2 .--- ------NOTE- -Not applicable to Reactor Trip Breaker Position.Perform CHANNEL CALIBRATION for each required 1 n instrumentation channel.Catawba Units 1 and 2 3.3.4-2 Amendment Nos4; , | SR 3.3.4.2 .--- ------NOTE- -Not applicable to Reactor Trip Breaker Position.Perform CHANNEL CALIBRATION for each required 1 n instrumentation channel.Catawba Units 1 and 2 3.3.4-2 Amendment Nos4; , | ||
LOP DG Start Instrumentation | LOP DG Start Instrumentation 3.3.5 SURVEILLANCE REQUIREMENTS FREQUENCY SURVEILLANCE SR 3.3.5.1-NOTE -------------------------------- | ||
REQUIREMENTS FREQUENCY SURVEILLANCE SR 3.3.5.1-NOTE -------------------------------- | |||
Testing shall consist of voltage sensor relay testing excluding actuation of load shedding diesel start, and time delay times.Perform TADOT.SR 3,3.5.2 Perform CHANNEL CALIBRATION with NOMINAL TRIP SETPOINT and Allowable Value as follows: a. Loss of voltage Allowable Value > 3242 V.Loss of voltage NOMINAL TRIP SETPOINT =3500 V.b. Degraded voltage Allowable Value > 3738 V.Degraded voltage NOMINAL TRIP SETPOINT 3766 V.Catawba Units 1 and 2 3-3.5-2 Amendment Nos.7 (n i~tt LOP DG Start Instrumentation B 3.3.5 BASES SURVEILLANCE REQUIREMENTS (continued) | Testing shall consist of voltage sensor relay testing excluding actuation of load shedding diesel start, and time delay times.Perform TADOT.SR 3,3.5.2 Perform CHANNEL CALIBRATION with NOMINAL TRIP SETPOINT and Allowable Value as follows: a. Loss of voltage Allowable Value > 3242 V.Loss of voltage NOMINAL TRIP SETPOINT =3500 V.b. Degraded voltage Allowable Value > 3738 V.Degraded voltage NOMINAL TRIP SETPOINT 3766 V.Catawba Units 1 and 2 3-3.5-2 Amendment Nos.7 (n i~tt LOP DG Start Instrumentation B 3.3.5 BASES SURVEILLANCE REQUIREMENTS (continued) | ||
SR 3.3.5.2 r SR 3.3.5.2 is the performance of a CHANNEL CALIBRATION. | SR 3.3.5.2 r SR 3.3.5.2 is the performance of a CHANNEL CALIBRATION. | ||
| Line 400: | Line 337: | ||
: 1. UFSAR, Section 8.3.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. UFSAR, Section 8.3.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3.3.5-6 Revision No. ba ... | Catawba Units 1 and 2 B 3.3.5-6 Revision No. ba ... | ||
Containment Air Release and Addition Isolation Instrumentation | Containment Air Release and Addition Isolation Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS | ||
REQUIREMENTS | |||
---------------------------- | ---------------------------- | ||
NOTE--------------------------------- | NOTE--------------------------------- | ||
| Line 436: | Line 369: | ||
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RHR or RCS loop is in operation. | SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RHR or RCS loop is in operation. | ||
SR 3.4.6.2 Verify SG secondary side water levels are > 12% narrow range for required RCS loops.2s$~&1 urs QjVs§rf SR 3.4.6.3 Verify correct breaker alignment and indicated power are available to the required pump that is not in operation. | SR 3.4.6.2 Verify SG secondary side water levels are > 12% narrow range for required RCS loops.2s$~&1 urs QjVs§rf SR 3.4.6.3 Verify correct breaker alignment and indicated power are available to the required pump that is not in operation. | ||
Catawba Units 1 and 2 3.4.6-2 Amendment Nos(ýýD | Catawba Units 1 and 2 3.4.6-2 Amendment Nos(ýýD | ||
[RCS I 1,(,,0-. -MAO)-S 5. Loops Filled 3.4-7 ACTIONlS CONDITION REQUIRED ACTION COMPLETION TIME-t *7 A. One RHR loop inoperable. | [RCS I 1,(,,0-. -MAO)-S 5. Loops Filled 3.4-7 ACTIONlS CONDITION REQUIRED ACTION COMPLETION TIME-t *7 A. One RHR loop inoperable. | ||
AND Required SGs secondary side water levels not within limits.A.1 Initiate action to restore a second RHR loop to OPERABLE status.OR A.2 Initiate action to restore required SG secondary side water levels to within limits.Immediately Immediately 4-B. Required RHR loops inoperable. | AND Required SGs secondary side water levels not within limits.A.1 Initiate action to restore a second RHR loop to OPERABLE status.OR A.2 Initiate action to restore required SG secondary side water levels to within limits.Immediately Immediately 4-B. Required RHR loops inoperable. | ||
| Line 449: | Line 382: | ||
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one RHR loop is in operation. | SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one RHR loop is in operation. | ||
SR 3.4.8.2 Verify correct breaker alignment and indicated power are available to the required RHR pump that is not in operation. | SR 3.4.8.2 Verify correct breaker alignment and indicated power are available to the required RHR pump that is not in operation. | ||
Catawba Units 1 and 2 3.4.8-2 Amendment Nos Pressurizer | Catawba Units 1 and 2 3.4.8-2 Amendment Nos Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water level is < 92% (1656 ft 3).E~2~SR 3.4.9.2 Verify capacity of each required group of pressurizer heaters is > 150 kW.SR 3.4.9.3 Verify required pressurizer heaters are capable of being powered from an emergency power supply.Catawba Units 1 and 2 3.4.9-2 Amendment Nos. (7731;?')\ýfý Pressurizer PORVs.3.4.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. (continued) | ||
REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water level is < 92% (1656 ft 3).E~2~SR 3.4.9.2 Verify capacity of each required group of pressurizer heaters is > 150 kW.SR 3.4.9.3 Verify required pressurizer heaters are capable of being powered from an emergency power supply.Catawba Units 1 and 2 3.4.9-2 Amendment Nos. (7731;?')\ýfý Pressurizer PORVs.3.4.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. (continued) | |||
F.2 Restore one block valve to 2 hours OPERABLE status if three block valves are inoperable. | F.2 Restore one block valve to 2 hours OPERABLE status if three block valves are inoperable. | ||
AND F.3 Restore remaining block 72 hours valve(s) to OPERABLE status.G. Required Action and G.1 Be in MODE 3. 6 hours associated Completion Time of Condition F not AND met.G.2 Be in MODE 4. 12 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1------------------------ | AND F.3 Restore remaining block 72 hours valve(s) to OPERABLE status.G. Required Action and G.1 Be in MODE 3. 6 hours associated Completion Time of Condition F not AND met.G.2 Be in MODE 4. 12 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1------------------------ | ||
| Line 462: | Line 391: | ||
Required to be performed in MODE 3 or MODE 4 when the temperature of all RCS cold legs is > 200'F.Perform a complete cycle of each PORV.SR 3.4.11.3------------------------------- | Required to be performed in MODE 3 or MODE 4 when the temperature of all RCS cold legs is > 200'F.Perform a complete cycle of each PORV.SR 3.4.11.3------------------------------- | ||
NO TE ----------------- | NO TE ----------------- | ||
This SR is not applicable to valve NC-36B.IAi&~EdZf | This SR is not applicable to valve NC-36B.IAi&~EdZf | ||
{Verify the nitrogen supply for each PORV is OPERABLE by: a. Manually transferring motive power from the air supply to the nitrogen supply, b. Isolating and venting the air supply, and c. Operating the PORV through one complete cycle.Catawba Units 1 and 2 3.4.11-4 Amendment Nos.( | {Verify the nitrogen supply for each PORV is OPERABLE by: a. Manually transferring motive power from the air supply to the nitrogen supply, b. Isolating and venting the air supply, and c. Operating the PORV through one complete cycle.Catawba Units 1 and 2 3.4.11-4 Amendment Nos.( | ||
LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of two pumps (charging, safety injection, or charging and safety injection) are capable of injecting into the RCS.SR 3.4.12.2 Verify each accumulator is isolated.SR 3.4.12.3 Verify RHR suction isolation valves are open for each required RHR suction relief valve.SR 3.4.12.4 Verify PORV block valve is open for each required r PORV.SR 3.4.12.5 ------ ..----------- | LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of two pumps (charging, safety injection, or charging and safety injection) are capable of injecting into the RCS.SR 3.4.12.2 Verify each accumulator is isolated.SR 3.4.12.3 Verify RHR suction isolation valves are open for each required RHR suction relief valve.SR 3.4.12.4 Verify PORV block valve is open for each required r PORV.SR 3.4.12.5 ------ ..----------- | ||
| Line 513: | Line 442: | ||
SURVEILLANCE FREQUENCY SR 3.6.3.8 Verify the combined leakage rate for all reactor building In accordance with bypass leakage paths is < 0.07 La when pressurized to > the Containment 14.68 psig. Leakage Rate Testing Program Catawba Units 1 and 2 3.6.3-7 Amendment | SURVEILLANCE FREQUENCY SR 3.6.3.8 Verify the combined leakage rate for all reactor building In accordance with bypass leakage paths is < 0.07 La when pressurized to > the Containment 14.68 psig. Leakage Rate Testing Program Catawba Units 1 and 2 3.6.3-7 Amendment | ||
.'os. 192/184 Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCO 3.6.4 APPLICABILITY: | .'os. 192/184 Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCO 3.6.4 APPLICABILITY: | ||
Containment pressure shall be > -0.1 psig and < +0.3 psig.MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment pressure A.1 Restore containment 1 hour not within limits, pressure to within limits.B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within limits.ý r <-i ýAn; tt: -Catawba Units 1 and 2-3.6.4-1 Amendment Nos.Oi Containment Air Temperature | Containment pressure shall be > -0.1 psig and < +0.3 psig.MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment pressure A.1 Restore containment 1 hour not within limits, pressure to within limits.B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within limits.ý r <-i ýAn; tt: -Catawba Units 1 and 2-3.6.4-1 Amendment Nos.Oi Containment Air Temperature 3.6.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment upper compartment average air temperature is within limits.SR 3.6.5.2 Verify containment lower compartment average air temperature is within limits.Catawba Units 1 and 2 3.6.5-2 Amendment Nos. (2ýý Containment Spray System 3-6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray System LCO 3.6.6 APPLICABILITY: | ||
REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment upper compartment average air temperature is within limits.SR 3.6.5.2 Verify containment lower compartment average air temperature is within limits.Catawba Units 1 and 2 3.6.5-2 Amendment Nos. (2ýý Containment Spray System 3-6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray System LCO 3.6.6 APPLICABILITY: | |||
Two containment spray trains shall be OPERABLE.MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One containment spray A-1 Restore containment spray 72 hours train inoperable, train to OPERABLE status.B. Required Action and B. 1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 5. 84 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each containment spray manual, power operated, and automatic valve in the flow pathl that is not locked, sealed, or otherwise secured in position is in the correct position.(continued) | Two containment spray trains shall be OPERABLE.MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One containment spray A-1 Restore containment spray 72 hours train inoperable, train to OPERABLE status.B. Required Action and B. 1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 5. 84 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each containment spray manual, power operated, and automatic valve in the flow pathl that is not locked, sealed, or otherwise secured in position is in the correct position.(continued) | ||
Catawba Units 1 and 2 3.6-6-1 Amendment Nos. G253 D248 Containment Spray System 3.6.6 SURVEILLANCE REQUIREMENTS (continued) | Catawba Units 1 and 2 3.6-6-1 Amendment Nos. G253 D248 Containment Spray System 3.6.6 SURVEILLANCE REQUIREMENTS (continued) | ||
| Line 677: | Line 602: | ||
NOTE --------Not applicable to DG batteries 24 months when battery has reached 85% of the expected life with capacity.> | NOTE --------Not applicable to DG batteries 24 months when battery has reached 85% of the expected life with capacity.> | ||
100% of manufacturer's rating Catawba Units 1 and 2 3.8.4-4 Amendment Nos.8 Batte'y Cell Parameters 3.8.6 t... IF,~ .'r-mI I A LIrSr rr-tI IIr~r-1. ar-I. I-I-t-~0uIr¶VrILL/MINL F I'r- Ur--IVlr'l I , SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters of the channels of DC and -DG batteries meet Table 3.8.6-1 Category A limits.SR 3.8.6.2 Not used.SR 3.8.6.3 Verify battery cell parameters of the channels of DC and DG batteries meet Table 3.8.6-1 Category B limits.AND Once within 7 days after a battery discharge<110 V AND Once within 7 days after a battery overcharge | 100% of manufacturer's rating Catawba Units 1 and 2 3.8.4-4 Amendment Nos.8 Batte'y Cell Parameters 3.8.6 t... IF,~ .'r-mI I A LIrSr rr-tI IIr~r-1. ar-I. I-I-t-~0uIr¶VrILL/MINL F I'r- Ur--IVlr'l I , SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters of the channels of DC and -DG batteries meet Table 3.8.6-1 Category A limits.SR 3.8.6.2 Not used.SR 3.8.6.3 Verify battery cell parameters of the channels of DC and DG batteries meet Table 3.8.6-1 Category B limits.AND Once within 7 days after a battery discharge<110 V AND Once within 7 days after a battery overcharge | ||
> 150 V-I SR 3.8.6.4 Verify average electrolyte temperature for the channels of DC and DG batteries of representative cells is > 60 0 F.Catawba Units 1 and 2 3.8.6-4 Amendment Nos e I10 CHANGES THIS PAGE."FOR INFORMATION ONLY Battery Cell Parameters | > 150 V-I SR 3.8.6.4 Verify average electrolyte temperature for the channels of DC and DG batteries of representative cells is > 60 0 F.Catawba Units 1 and 2 3.8.6-4 Amendment Nos e I10 CHANGES THIS PAGE."FOR INFORMATION ONLY Battery Cell Parameters 3.8.6 Table 3.8.6-1 (page 1 of 1)Battery Cell Parameters Requirements CATEGORY A: CATEGORY C: LIMITS FOR EACH CATEGORY B: ALLOWABLE DESIGNATED LIMITS FOR EACH LIMITS FOR EACH PARAMETER PILOT CELL CONNECTED CELL CONNECTED CELL Electrolyte Level > Minimum level > Minimum level Above top of plates, indication mark, and indication mark, and and not overflowing | ||
3.8.6-1 (page 1 of 1)Battery Cell Parameters Requirements CATEGORY A: CATEGORY C: LIMITS FOR EACH CATEGORY B: ALLOWABLE DESIGNATED LIMITS FOR EACH LIMITS FOR EACH PARAMETER PILOT CELL CONNECTED CELL CONNECTED CELL Electrolyte Level > Minimum level > Minimum level Above top of plates, indication mark, and indication mark, and and not overflowing | |||
<Y4 inch above < 1/4 inch above maximum level maximum level indication mark(a) indication mark(a)Float Voltage > 2.13 V > 2.13 V > 2.07 V Specific Gravity(b)(c) | <Y4 inch above < 1/4 inch above maximum level maximum level indication mark(a) indication mark(a)Float Voltage > 2.13 V > 2.13 V > 2.07 V Specific Gravity(b)(c) | ||
> 1.200 > 1.195 Not more than 0.020 below average of all AND connected cells or> 1.195 Average of all connected cells AND> 1.205 Average of all connected cells> 1.195 (a) It is acceptable for the electrolyte level to temporarily increase above the specified maximum during equalizing charges provided it is not overflowing.(b) Corrected for electrolyte temperature and level. Level correction is not required, however, when battery charging is < 2 amps when on float charge.(c) A battery charging current of < 2 amps when on float charge is acceptable for meeting specific gravity limits following a battery recharge, for a maximum of 7 days. When charging current is used to satisfy specific gravity requirements, specific gravity of each connected cell shall be measured prior to expiration of the 7 day allowance. | > 1.200 > 1.195 Not more than 0.020 below average of all AND connected cells or> 1.195 Average of all connected cells AND> 1.205 Average of all connected cells> 1.195 (a) It is acceptable for the electrolyte level to temporarily increase above the specified maximum during equalizing charges provided it is not overflowing.(b) Corrected for electrolyte temperature and level. Level correction is not required, however, when battery charging is < 2 amps when on float charge.(c) A battery charging current of < 2 amps when on float charge is acceptable for meeting specific gravity limits following a battery recharge, for a maximum of 7 days. When charging current is used to satisfy specific gravity requirements, specific gravity of each connected cell shall be measured prior to expiration of the 7 day allowance. | ||
| Line 694: | Line 616: | ||
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to required AC, DC channel, DC train, and AC vital bus electrical power distribution subsystems. | SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to required AC, DC channel, DC train, and AC vital bus electrical power distribution subsystems. | ||
G3 ýJ ___________________ | G3 ýJ ___________________ | ||
Catawba Units 1 and 2 3.8.10-2 Amendment No1165 Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS | Catawba Units 1 and 2 3.8.10-2 Amendment No1165 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 rofueling canal, and the refueling cavity shall be maintained within the limit I : fied in the COLR.--...-.-.- | ||
Concentration LCO 3.9.1 Boron concentrations of the Reactor Coolant System, the rofueling canal, and the refueling cavity shall be maintained within the limit I : fied in the COLR.--...-.-.- | |||
..------------------- | ..------------------- | ||
NOTE------- | NOTE------- | ||
| Line 704: | Line 623: | ||
AND A.2 Suspend positive reactivity Immediately additions. | AND A.2 Suspend positive reactivity Immediately additions. | ||
AND A.3 Initiate action to restore Immediately boron concentration to within limit.Catawba Units 1 and 2 3.9.1-1 Amendment Nos. | AND A.3 Initiate action to restore Immediately boron concentration to within limit.Catawba Units 1 and 2 3.9.1-1 Amendment Nos. | ||
Nuclear Instrumentation | Nuclear Instrumentation 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4-SR 3.9.2.1 Perform CHANNEL CHECK.SR 3.9.2.2-NO T E --------------------------------- | ||
REQUIREMENTS SURVEILLANCE FREQUENCY 4-SR 3.9.2.1 Perform CHANNEL CHECK.SR 3.9.2.2-NO T E --------------------------------- | |||
Neutron detectors are excluded from CHANNEL CALIBRATION. | Neutron detectors are excluded from CHANNEL CALIBRATION. | ||
ON .ýdrxtl-Perform CHANNEL CALIBRATION. | ON .ýdrxtl-Perform CHANNEL CALIBRATION. | ||
Catawba Units 1 and 2 3.9.2-2 Amendment Nos. (1 257ý20, Containment Penetrations | Catawba Units 1 and 2 3.9.2-2 Amendment Nos. (1 257ý20, Containment Penetrations 3.9.3 ACTIONS (continued) | ||
(continued) | |||
CONDITION REQUIRED ACTION COMPLETION TIME B. One or more CPES B.1 Restore CPES train(s) 7 days train(s) heater heater to OPERABLE inoperable, status.OR B.2 Initiate action in 7 days accordance with Specification 5.6.6.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify each required containment penetration is in the required status.SR 3.9.3.2 Operate each CPES for > 10 continuous hours with the heaters operating. | CONDITION REQUIRED ACTION COMPLETION TIME B. One or more CPES B.1 Restore CPES train(s) 7 days train(s) heater heater to OPERABLE inoperable, status.OR B.2 Initiate action in 7 days accordance with Specification 5.6.6.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify each required containment penetration is in the required status.SR 3.9.3.2 Operate each CPES for > 10 continuous hours with the heaters operating. | ||
SR 3.9.3.3 Perform required CPES filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). the VFTP 0 Catawba Units I and 2 3.9.3-2 Amendment Nos (173//65 RHR and Coolant Circulation | SR 3.9.3.3 Perform required CPES filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). the VFTP 0 Catawba Units I and 2 3.9.3-2 Amendment Nos (173//65 RHR and Coolant Circulation | ||
| Line 725: | Line 637: | ||
SURVEILLANCE REQUIREMENTS SURVEILLANCE IFREQUENCY SR 3.9.5.1 Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of > 1000 gpm and RCS temperature is < 140'F.SR 3.9.5.2 Verify correct breaker alignment and indicated power available to the required RHR pump that is not in operation. | SURVEILLANCE REQUIREMENTS SURVEILLANCE IFREQUENCY SR 3.9.5.1 Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of > 1000 gpm and RCS temperature is < 140'F.SR 3.9.5.2 Verify correct breaker alignment and indicated power available to the required RHR pump that is not in operation. | ||
Catawba Units 1 and 2 3.9.5-2 Amendment Nos. 0173/!ý:) | Catawba Units 1 and 2 3.9.5-2 Amendment Nos. 0173/!ý:) | ||
Refueling Cavity Water Level 3.9.6 3.9 REFUELING OPERATIONS | Refueling Cavity Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Refueling Cavity Water Level LCO 3.9.6 APPLICABILITY: | ||
Cavity Water Level LCO 3.9.6 APPLICABILITY: | |||
Refueling cavity water level shall be maintained | Refueling cavity water level shall be maintained | ||
> 23 ft above the top of reactor vessel flange.During CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, During movement of irradiated fuel assemblies within containment. | > 23 ft above the top of reactor vessel flange.During CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, During movement of irradiated fuel assemblies within containment. | ||
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refueling cavity water A.1 Suspend CORE Immediately level not within limit. ALTERATIONS. | 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. | 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 top of reactor vessel flange.K= -Catawba Units 1 and 2 3.9.6-1 Amendment Unborated Water Source Isolation Valves 3.9.7 3.9 REFUELING OPERATIONS | SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify refueling cavity water level is > 23 ft above the top of reactor vessel flange.K= -Catawba Units 1 and 2 3.9.6-1 Amendment Unborated Water Source Isolation Valves 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Unborated Water Source Isolation Valves LCO 3.9.7 Each valve used to isolate unborated water sources shall be secured in the closed position.APPLICABILITY: | ||
Water Source Isolation Valves LCO 3.9.7 Each valve used to isolate unborated water sources shall be secured in the closed position.APPLICABILITY: | |||
MODE 6.ACTIONS---------------------------------------------------- | MODE 6.ACTIONS---------------------------------------------------- | ||
I\ 1t j r ----------------------------------- | I\ 1t j r ----------------------------------- | ||
| Line 756: | Line 660: | ||
: d. Fuel burnup based on gross thermal energy generation; | : d. Fuel burnup based on gross thermal energy generation; | ||
: e. Xenon concentration; | : e. Xenon concentration; | ||
: f. Samarium concentration; and g. Isothermal temperature coefficient (ITC).Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.hFre enc of 24 ours' base on the enera slo cha e in r quire bor con ntrati n an he low robabacent occu ing thout e re ired M. T sallo tim for t oper or to.co1 ct t requi ed da ,wi inclu s pe rmin *a boon 'nce ration nalysj , an_ ompi e the alcul on.REFERENCES | : f. Samarium concentration; and g. Isothermal temperature coefficient (ITC).Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.hFre enc of 24 ours' base on the enera slo cha e in r quire bor con ntrati n an he low robabacent occu ing thout e re ired M. T sallo tim for t oper or to.co1 ct t requi ed da ,wi inclu s pe rmin *a boon 'nce ration nalysj , an_ ompi e the alcul on.REFERENCES | ||
: 1. 10 CFR 50, Appendix A, GDC 26.2. UFSAR, Section 15.1.5.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. 10 CFR 50, Appendix A, GDC 26.2. UFSAR, Section 15.1.5.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
: 4. UFSAR, Section 15.4.6.5. 10 CFR 50.67.Catawba Units 1 and 2 B 3.1.1-6 Revision No.(R Core Reactivity B 3.1.2 BASES ACTIONS (continued) | : 4. UFSAR, Section 15.4.6.5. 10 CFR 50.67.Catawba Units 1 and 2 B 3.1.1-6 Revision No.(R Core Reactivity B 3.1.2 BASES ACTIONS (continued) | ||
B.1 If the core reactivity cannot be restored to within the 1 % Ak/k limit, 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 hours. If the SDM for MODE 3 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 3 from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations. | B.1 If the core reactivity cannot be restored to within the 1 % Ak/k limit, 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 hours. If the SDM for MODE 3 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 3 from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE SR 3.1.2.1 REQUIREMENTS 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. | 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. | ||
The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization of predicted core reactivity to the measured value must take place within the first 60 effective full power days (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 __esi n calculat s. Te re ire ub e ue Freq ncy 31 FP isac pt e, sed nct sl cha ges ue t fuel eple i n n thpre nc f o in~(T AFD etc. forp mp ndi itio f a no al REFERENCES | The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization of predicted core reactivity to the measured value must take place within the first 60 effective full power days (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 __esi n calculat s. Te re ire ub e ue Freq ncy 31 FP isac pt e, sed nct sl cha ges ue t fuel eple i n n thpre nc f o in~(T AFD etc. forp mp ndi itio f a no al REFERENCES | ||
: 1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specification, (c)(2)(ii). | : 1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specification, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3.1.2-5 Revision No-@ | Catawba Units 1 and 2 B 3.1.2-5 Revision No-@ | ||
| Line 818: | Line 722: | ||
If FMm(X,Y) is evaluated and found to be within its surveillance limit, an evaluation is required to account for any increase to FM (X,Y) that may occur and cause the FM(X,Y)suRv limit to be exceeded before the next required FH(X,Y)SuRv evaluation. | If FMm(X,Y) is evaluated and found to be within its surveillance limit, an evaluation is required to account for any increase to FM (X,Y) that may occur and cause the FM(X,Y)suRv limit to be exceeded before the next required FH(X,Y)SuRv evaluation. | ||
In addition to ensuring via surveillance that the enthalpy rise hot channel factor is within its steady state and surveillance limits when a measurement is taken, there are also requirements to extrapolate trends in both the measured hot channel factor and in its surveillance limit. Two extrapolations are performed for this limit: Catawba Units 1 and 2 B 3.2.2-8 Revision NoZ,ý FAH(X,Y)B 3.2.2 BASES SURVEILLANCE REQUIREMENTS (continued) | In addition to ensuring via surveillance that the enthalpy rise hot channel factor is within its steady state and surveillance limits when a measurement is taken, there are also requirements to extrapolate trends in both the measured hot channel factor and in its surveillance limit. Two extrapolations are performed for this limit: Catawba Units 1 and 2 B 3.2.2-8 Revision NoZ,ý FAH(X,Y)B 3.2.2 BASES SURVEILLANCE REQUIREMENTS (continued) | ||
: 1. The first extrapolation determines whether the measured enthalpy rise hot channel factor is likely to exceed its surveillance limit prior to the next performance of the SR.2. The second extrapolation determines whether, prior to the next performance of the SR, the ratio of the measured enthalpy rise hot channel factor to the surveillance limit is likely to decrease below the value of that ratio when the measurement was taken.Each of these extrapolations is applied separately to the enthalpy rise hot channel factor surveillance limit. If both of the extrapolations are unfavorable, i.e., if the extrapolated factor is expected to exceed the extrapolated limit and the extrapolated factor is expected to become a larger fraction of the extrapolated limit than the measured factor is of the current limit, additional actions must be taken. These actions are to meet the FMAH(X,Y) limit with the last FMAH(X,Y) increased by a factor of 1.02, or to evaluate FMAH(X,Y) prior to the point in time when the extrapolated values are expected to exceed the extrapolated limits. These alternative requirements attempt to prevent FMAH(X,Y) from exceeding its limit for any significant period of time without detection using the best available data.F MAH(X,Y) is not required to be extrapolated for the initial flux map taken after reaching equilibrium conditions since the initial flux map establishes the baseline measurement for future trending.FMAH(X,Y) is verified at power levels 10% RTP above the THERMAL POWER of its last verification, 12 hours after achieving equilibrium conditions to ensure that FMAH(X,Y) is within its limit at high power levels.The Surve)f'ance Freq ncy of 31 E D is adeq te to monitor change 9 power distriution with c/re burnup. 7he Surveillanc may be done m re freauentl4 if reauired bv the results f FMAW(X.Y) eva uations.REFERENCES | : 1. The first extrapolation determines whether the measured enthalpy rise hot channel factor is likely to exceed its surveillance limit prior to the next performance of the SR.2. The second extrapolation determines whether, prior to the next performance of the SR, the ratio of the measured enthalpy rise hot channel factor to the surveillance limit is likely to decrease below the value of that ratio when the measurement was taken.Each of these extrapolations is applied separately to the enthalpy rise hot channel factor surveillance limit. If both of the extrapolations are unfavorable, i.e., if the extrapolated factor is expected to exceed the extrapolated limit and the extrapolated factor is expected to become a larger fraction of the extrapolated limit than the measured factor is of the current limit, additional actions must be taken. These actions are to meet the FMAH(X,Y) limit with the last FMAH(X,Y) increased by a factor of 1.02, or to evaluate FMAH(X,Y) prior to the point in time when the extrapolated values are expected to exceed the extrapolated limits. These alternative requirements attempt to prevent FMAH(X,Y) from exceeding its limit for any significant period of time without detection using the best available data.F MAH(X,Y) is not required to be extrapolated for the initial flux map taken after reaching equilibrium conditions since the initial flux map establishes the baseline measurement for future trending.FMAH(X,Y) is verified at power levels 10% RTP above the THERMAL POWER of its last verification, 12 hours after achieving equilibrium conditions to ensure that FMAH(X,Y) is within its limit at high power levels.The Surve)f'ance Freq ncy of 31 E D is adeq te to monitor change 9 power distriution with c/re burnup. 7he Surveillanc may be done m re freauentl4 if reauired bv the results f FMAW(X.Y) eva uations.REFERENCES | ||
: 1. UFSAR Section 15.4.8 2. 10 CFR 50, Appendix A, GDC 26.Catawba Units 1 and 2 B 3.2.2-9 Revision Noo NO CHANGES THIS PAGE.FOR INFORMATION ONLY AFD B 3.2.3 BASES LCO (continued) | : 1. UFSAR Section 15.4.8 2. 10 CFR 50, Appendix A, GDC 26.Catawba Units 1 and 2 B 3.2.2-9 Revision Noo NO CHANGES THIS PAGE.FOR INFORMATION ONLY AFD B 3.2.3 BASES LCO (continued) | ||
Signals are available to the operator from the Nuclear Instrumentation System (NIS) excore neutron detectors (Ref. 3). Separate signals are taken from the top and bottom detectors. | Signals are available to the operator from the Nuclear Instrumentation System (NIS) excore neutron detectors (Ref. 3). Separate signals are taken from the top and bottom detectors. | ||
| Line 839: | Line 743: | ||
The two sets of four symmetric thimbles is a set of eight unique detector locations. | The two sets of four symmetric thimbles is a set of eight unique detector locations. | ||
These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.The symmetric thimble flux map can be used to generate symmetric thimble "tilt." This can be compared to a reference symmetric thimble tilt, from the most recent full core flux map, to generate an incore tilt.Therefore, incore tilt can be used to confirm that QPTR is within limits.With one or more NIS channel inputs to QPTR inoperable, the indicated tilt may be changed from the value indicated with all four channels OPERABLE. | These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.The symmetric thimble flux map can be used to generate symmetric thimble "tilt." This can be compared to a reference symmetric thimble tilt, from the most recent full core flux map, to generate an incore tilt.Therefore, incore tilt can be used to confirm that QPTR is within limits.With one or more NIS channel inputs to QPTR inoperable, the indicated tilt may be changed from the value indicated with all four channels OPERABLE. | ||
To confirm that no change in tilt has actually occurred, Catawba Units 1 and 2 B 3.2.4-6 Revision No[D NO CHANGES THIS PAGE.FOR INFORMATION ONLY QPTR~B 3.2.4 BASES i_ __.....SURVEILLANCE REQUIREMENTS (continued) which might cause the QPTR limit to be exceeded, the incore result may be compared against previous flux maps either using the symmetric thimbles as described above or a complete flux map. Nominally, quadrant tilt from the Surveillance should be within 2% of the tilt shown by the most recent flux map data.REFERENCES | To confirm that no change in tilt has actually occurred, Catawba Units 1 and 2 B 3.2.4-6 Revision No[D NO CHANGES THIS PAGE.FOR INFORMATION ONLY QPTR~B 3.2.4 BASES i_ __.....SURVEILLANCE REQUIREMENTS (continued) which might cause the QPTR limit to be exceeded, the incore result may be compared against previous flux maps either using the symmetric thimbles as described above or a complete flux map. Nominally, quadrant tilt from the Surveillance should be within 2% of the tilt shown by the most recent flux map data.REFERENCES | ||
: 1. 10 CFR 50.46.2. UFSAR Section 15.4.8.3. 10 CFR 50, Appendix A, GDC 26.4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. 10 CFR 50.46.2. UFSAR Section 15.4.8.3. 10 CFR 50, Appendix A, GDC 26.4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3.2.4-7 Revision No. 1 RTS Instrumentation B 3.3.1 BASES ACTIONS (continued) | Catawba Units 1 and 2 B 3.2.4-7 Revision No. 1 RTS Instrumentation B 3.3.1 BASES ACTIONS (continued) | ||
| Line 898: | Line 802: | ||
cceptable fr~t a reiblt | cceptable fr~t a reiblt | ||
,_._ .. .SR 3.3.1.16 is modified by a Note stating that neutron deiedt-br6, are, excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response. | ,_._ .. .SR 3.3.1.16 is modified by a Note stating that neutron deiedt-br6, are, excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response. | ||
The response time of the neutron flux signal portion of the channel shall be measured from detector output or input of the first electronic component in the channel.REFERENCES | The response time of the neutron flux signal portion of the channel shall be measured from detector output or input of the first electronic component in the channel.REFERENCES | ||
: 1. UFSAR, Chapter 7.2. UFSAR, Chapter 6.3. UFSAR, Chapter 15.4. IEEE-279-1971. | : 1. UFSAR, Chapter 7.2. UFSAR, Chapter 6.3. UFSAR, Chapter 15.4. IEEE-279-1971. | ||
: 5. 10 CFR 50.49.6. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 5. 10 CFR 50.49.6. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
| Line 946: | Line 850: | ||
SR 3.3.3.3ý C LC LBRATI is perfogm nl r a r.imately -at eve _efu CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy. | SR 3.3.3.3ý C LC LBRATI is perfogm nl r a r.imately -at eve _efu CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy. | ||
This SR is modified by two Notes. Note 1 excludes neutron detectors. | This SR is modified by two Notes. Note 1 excludes neutron detectors. | ||
The calibration method for neutron detectors is specified in the Bases of LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." Note 2 describes the calibration methods for.the Containment Area -High Range monitor. phperrequFicys bases4K, REFERENCES | The calibration method for neutron detectors is specified in the Bases of LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." Note 2 describes the calibration methods for.the Containment Area -High Range monitor. phperrequFicys bases4K, REFERENCES | ||
: 1. UFSAR Section 1.8.2. Regulatory Guide 1.97, Rev. 2. , 3. NUREG-0737, Supplement 1, "TMI Action Items." 4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. UFSAR Section 1.8.2. Regulatory Guide 1.97, Rev. 2. , 3. NUREG-0737, Supplement 1, "TMI Action Items." 4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3.3.3-16 Revision No.0 Remote Shutdown System B 3.3.4 BASES ACTIONS (continued) | Catawba Units 1 and 2 B 3.3.3-16 Revision No.0 Remote Shutdown System B 3.3.4 BASES ACTIONS (continued) | ||
| Line 959: | Line 863: | ||
Catawba Units 1 and 2 B 3.3.4-4 Revision No.O Remote Shutdown System B 3.3.4 BASES SURVEILLANCE REQUIREMENTS (continued) | Catawba Units 1 and 2 B 3.3.4-4 Revision No.O Remote Shutdown System B 3.3.4 BASES SURVEILLANCE REQUIREMENTS (continued) | ||
SR 3.3.4.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within' the necessary range and accuracy.The surveillance is modified by a Note that excepts the RTB position indication from a CHANNEL CALIBRATION. | SR 3.3.4.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within' the necessary range and accuracy.The surveillance is modified by a Note that excepts the RTB position indication from a CHANNEL CALIBRATION. | ||
The RTB position is indicated by a mechanical "flag" on the breaker.REFERENCES | The RTB position is indicated by a mechanical "flag" on the breaker.REFERENCES | ||
: 1. 10 CFR 50, Appendix A, GDC 19.Catawba Units 1 and 2 B 3.3.4-5 Revision No.0 LOP DG Start Instrumentation B 3.3.5 BASES ACTIONS (continued) | : 1. 10 CFR 50, Appendix A, GDC 19.Catawba Units 1 and 2 B 3.3.4-5 Revision No.0 LOP DG Start Instrumentation B 3.3.5 BASES ACTIONS (continued) | ||
B.1 Condition B applies when more than one loss of voltage or more than one degraded voltage channel on a single bus is inoperable. | B.1 Condition B applies when more than one loss of voltage or more than one degraded voltage channel on a single bus is inoperable. | ||
| Line 980: | Line 884: | ||
:-u 1)WO pAP-1390 , 0. For slave r s or any auxiliary relays e ircuit that are of the type of SlaeReay Westingh seAR or Potter & Brumfie MDR" the SLAVE RELAY TES Test Interval April 1994;2) is perfor ed every 18 months. This est frequeno is based on the r 2 WCAP-1 7 Revision 2-P-A, reliabili assessments presented* | :-u 1)WO pAP-1390 , 0. For slave r s or any auxiliary relays e ircuit that are of the type of SlaeReay Westingh seAR or Potter & Brumfie MDR" the SLAVE RELAY TES Test Interval April 1994;2) is perfor ed every 18 months. This est frequeno is based on the r 2 WCAP-1 7 Revision 2-P-A, reliabili assessments presented* | ||
Referencehese"Relia ity Assessment of reliabs ty assessments are relay nd apply only to the s tiaose Type AR Relay Ws anghouse Type AR Relay .e inghouse AR an r rumfield MDR t e relays. SS S sla\Agst 3) WCAP- 878- rel s o xiliary relay not addressed b d not qua P-A Revision 2, "Relit ility extended surveillance' tervals and will continue to bet ted a 9, ro, r 5 Ff- r Awd" A'a -P, --,,ý /tA I/SR 3.3.6.4 WCAP-13900 ension of Slave Relay ellance Test Int. s," April 1994, For slave relays or any auxiliary relays in the circuit that are of the type Westinghouse AR or Potter &Brumfield MDR, the SLAVE RELAY TEST frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.SR 3.3.6.4 is the performance of a TADOT. This test is a check of the Manual Actuation Functions and is performed every 18 months. Each Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, etc.).The test also includes trip devices that provide actuation signals directly to the SSPS, bypassing the analog process control equipment. | Referencehese"Relia ity Assessment of reliabs ty assessments are relay nd apply only to the s tiaose Type AR Relay Ws anghouse Type AR Relay .e inghouse AR an r rumfield MDR t e relays. SS S sla\Agst 3) WCAP- 878- rel s o xiliary relay not addressed b d not qua P-A Revision 2, "Relit ility extended surveillance' tervals and will continue to bet ted a 9, ro, r 5 Ff- r Awd" A'a -P, --,,ý /tA I/SR 3.3.6.4 WCAP-13900 ension of Slave Relay ellance Test Int. s," April 1994, For slave relays or any auxiliary relays in the circuit that are of the type Westinghouse AR or Potter &Brumfield MDR, the SLAVE RELAY TEST frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.SR 3.3.6.4 is the performance of a TADOT. This test is a check of the Manual Actuation Functions and is performed every 18 months. Each Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, etc.).The test also includes trip devices that provide actuation signals directly to the SSPS, bypassing the analog process control equipment. | ||
The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.The Frequen, is ba on the known reliab -of the Functior r ancy avai le, and has been sln to be acceptaJ:ef operating e rience. /REFERENCES | The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.The Frequen, is ba on the known reliab -of the Functior r ancy avai le, and has been sln to be acceptaJ:ef operating e rience. /REFERENCES | ||
: 1. 10 CFR 50.67.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. 10 CFR 50.67.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
WCAP-13900, ension of Slave Relay Surveillance | WCAP-13900, ension of Slave Relay Surveillance | ||
| Line 1,014: | Line 918: | ||
.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. | .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. | 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 | REFERENCES | ||
: 1. 10 CFR 50, Appendix G.2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.3. ASTM E 185-73, 1973 (Unit 1), E 185-82, 1982 (Unit 2).4. 10 CFR 50, Appendix H.5. Regulatory Guide 1.99, Revision 2, May 1988.6. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E.7. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. 10 CFR 50, Appendix G.2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.3. ASTM E 185-73, 1973 (Unit 1), E 185-82, 1982 (Unit 2).4. 10 CFR 50, Appendix H.5. Regulatory Guide 1.99, Revision 2, May 1988.6. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E.7. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3.4.3-6 Revision No-RCS Loops -MODES 1 and 2 B 3.4.4 BASES APPLICABILITY (continued) | Catawba Units 1 and 2 B 3.4.3-6 Revision No-RCS Loops -MODES 1 and 2 B 3.4.4 BASES APPLICABILITY (continued) | ||
LCO 3.4.5, "RCS Loops-MODE 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 3.4.17, "RCS Loops-Test Exceptions"; | LCO 3.4.5, "RCS Loops-MODE 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 3.4.17, "RCS Loops-Test Exceptions"; | ||
LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (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 3. 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 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verificationC IjIithat each RCS loop is in operation. | LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (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 3. 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 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verificationC IjIithat each RCS loop is in operation. | ||
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is rovidinc val wleaintainin th ar to Freque of is suffiienle siderin er indicatioand alar avail e to ope or in the c rol room to m itor RCS op perfrmanc REFERENCES | Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is rovidinc val wleaintainin th ar to Freque of is suffiienle siderin er indicatioand alar avail e to ope or in the c rol room to m itor RCS op perfrmanc REFERENCES | ||
: 1. UFSAR, Section 15.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. UFSAR, Section 15.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3.4.4-3 Revision RCS Loops -MODE 3 B 3.4.5 BASES ACTIONS (continued) | Catawba Units 1 and 2 B 3.4.4-3 Revision RCS Loops -MODE 3 B 3.4.5 BASES ACTIONS (continued) | ||
| Line 1,048: | Line 952: | ||
SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is> 12%. If the SG secondary side narrow range water level is < 12%, the tubes may become uncovered and the associated loop may not be ca able of providing the heat sink necessary for removal of decay hg.t-T 2h ureuncy is consid .ed adequa e Fin ew of other Catwicati°U con1ro ad 2e art Be :4ra.t r t os Catawba Units 1 and 2 B 3.4.6-4 Revision Nol" RCS Loops -MODE 4 B 3.4.6 BASES SURVEILLANCE REQUIREMENTS (continued) | SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is> 12%. If the SG secondary side narrow range water level is < 12%, the tubes may become uncovered and the associated loop may not be ca able of providing the heat sink necessary for removal of decay hg.t-T 2h ureuncy is consid .ed adequa e Fin ew of other Catwicati°U con1ro ad 2e art Be :4ra.t r t os Catawba Units 1 and 2 B 3.4.6-4 Revision Nol" RCS Loops -MODE 4 B 3.4.6 BASES SURVEILLANCE REQUIREMENTS (continued) | ||
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. | 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 ver in--roper breaker alignment and ower a.ailable to the re uired pump..The irýquency of.3ys is co idered reonabi view of ramns acontroe a has bee -h bown REFERENCES | Verification is performed by ver in--roper breaker alignment and ower a.ailable to the re uired pump..The irýquency of.3ys is co idered reonabi view of ramns acontroe a has bee -h bown REFERENCES | ||
: 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3.4.6-5 Revision Noo RCS Loops -MODE 5. Loops Filled B 34.7 BASES ACTIONS A.1 and A.2 If one RHR loop is inoperable and the required SGs have secondary side narrow range water levels < 12%, redundancy for heat removal is lost Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore the required SG secondary side water levels. Either Required Action A.1 or Required Action A.2 will restore redundant heat removal paths. The immediate Completion Time reflects.the importance of maintaining the availability of two paths for heat removal.B.1 and B.2 If no RHR loop is in operation, except during conditions permitted by Note 1, or if no loop is OPERABLE,.all operations involving introduction of coolant into the ROS with boron concentration less than required to meet SDM of LCO 3.1.1 must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated. | Catawba Units 1 and 2 B 3.4.6-5 Revision Noo RCS Loops -MODE 5. Loops Filled B 34.7 BASES ACTIONS A.1 and A.2 If one RHR loop is inoperable and the required SGs have secondary side narrow range water levels < 12%, redundancy for heat removal is lost Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore the required SG secondary side water levels. Either Required Action A.1 or Required Action A.2 will restore redundant heat removal paths. The immediate Completion Time reflects.the importance of maintaining the availability of two paths for heat removal.B.1 and B.2 If no RHR loop is in operation, except during conditions permitted by Note 1, or if no loop is OPERABLE,.all operations involving introduction of coolant into the ROS with boron concentration less than required to meet SDM of LCO 3.1.1 must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated. | ||
| Line 1,057: | Line 961: | ||
Verification includes I1ow-r-af-e, temperature, or pump status monitorin , which help ensure that forced flow s vn at r oval.'e Fr- e'nc--y 12 o o ufficient dering rdi ictin -nd ala availa orator in -mcotrolmrortomnt4Rlop,SR 3.4.7.2 Verifying that at least two SGs are OPERABLE by ensuring their secondary side narrow ranqe water levels are >_ 12% en 4 i!res an tqfrr'!-decay heat removal method in the event that the second RHR loop is not OPERAB th RHR loos are OPERABLE, this Surveillance is not needed The o -Freque is co edred equate iaew of o r din lleons ieinth antro, in to al the ooe or toth ssof Catawba Units 1 and 2 B 3.4.7-4 Revision No(2 RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued) | Verification includes I1ow-r-af-e, temperature, or pump status monitorin , which help ensure that forced flow s vn at r oval.'e Fr- e'nc--y 12 o o ufficient dering rdi ictin -nd ala availa orator in -mcotrolmrortomnt4Rlop,SR 3.4.7.2 Verifying that at least two SGs are OPERABLE by ensuring their secondary side narrow ranqe water levels are >_ 12% en 4 i!res an tqfrr'!-decay heat removal method in the event that the second RHR loop is not OPERAB th RHR loos are OPERABLE, this Surveillance is not needed The o -Freque is co edred equate iaew of o r din lleons ieinth antro, in to al the ooe or toth ssof Catawba Units 1 and 2 B 3.4.7-4 Revision No(2 RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued) | ||
SR 3.4.7.3 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. | SR 3.4.7.3 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 RHR pump.If secondary side narrow range water level is > 12% in at least two SGs, this Surveillance is not needed. Th F uency of 7a is consi reasonable in- v of o er admini tive as been.shown to Jaeacceptable by op ti ng e REFERENCES | Verification is performed by verifying proper breaker alignment and power available to the RHR pump.If secondary side narrow range water level is > 12% in at least two SGs, this Surveillance is not needed. Th F uency of 7a is consi reasonable in- v of o er admini tive as been.shown to Jaeacceptable by op ti ng e REFERENCES | ||
: 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3.4.7-5/A Revision Noo RCS Loops -MODE 5, Loops Not Filled B 3.4.8 BASES ACTIONS (continued) immediately to restore an RHR loop to OPERABLE status and operation. | Catawba Units 1 and 2 B 3.4.7-5/A Revision Noo RCS Loops -MODE 5, Loops Not Filled B 3.4.8 BASES ACTIONS (continued) immediately to restore an RHR loop to OPERABLE status and operation. | ||
| Line 1,067: | Line 971: | ||
SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification 12 urs that one loop is in operation. | SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification 12 urs that one loop is in operation. | ||
Verification includes flow rate, temperture, or pump status monitoring, which hel ensure that forced flow is providing heat removal. be reqenc f 112:ý hoF r ( suff ; .t sidTe-r, n-g`-omURrndicatio'n .nd!alarms vail control rX fnto monit H'R l-- oop.SR 3.4.8.2 Verification that the required number of pumps are OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. | Verification includes flow rate, temperture, or pump status monitoring, which hel ensure that forced flow is providing heat removal. be reqenc f 112:ý hoF r ( suff ; .t sidTe-r, n-g`-omURrndicatio'n .nd!alarms vail control rX fnto monit H'R l-- oop.SR 3.4.8.2 Verification that the required number of pumps are 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 ro per breaker alignment and power availabla-, _t required pumps., he Frouencyo days is clsidered r asonajie in view of oter administratiWe contr(il abl nd has be6n be acc able by operatio experihnc REFERENCES | Verification is performed by verifying ro per breaker alignment and power availabla-, _t required pumps., he Frouencyo days is clsidered r asonajie in view of oter administratiWe contr(il abl nd has be6n be acc able by operatio experihnc REFERENCES | ||
: 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3.4.8-3 Revis on No Pressurizer B 3.4.9 BASES ACTIONS (continued) | Catawba Units 1 and 2 B 3.4.8-3 Revis on No Pressurizer B 3.4.9 BASES ACTIONS (continued) | ||
| Line 1,104: | Line 1,008: | ||
During normal operation the primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling.e 72 hour equency is a reasona e-lnterval to trend pr ary to second EAKAGE and reco gades the importanc early leak e det ion in th reventior eaccidents and redu ion of ote nsequences. | During normal operation the primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling.e 72 hour equency is a reasona e-lnterval to trend pr ary to second EAKAGE and reco gades the importanc early leak e det ion in th reventior eaccidents and redu ion of ote nsequences. | ||
Note und6r the Frequency column states that this SR is only require to e.perfo duxa,,steady state operation. | Note und6r the Frequency column states that this SR is only require to e.perfo duxa,,steady state operation. | ||
REFERENCES | REFERENCES | ||
: 1. 10 CFR 50, Appendix A, GDC 30.2. Regulatory Guide 1 .45, May 1973.3. UFSAR, Section 15.4. 1-&-CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. 10 CFR 50, Appendix A, GDC 30.2. Regulatory Guide 1 .45, May 1973.3. UFSAR, Section 15.4. 1-&-CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
: 5. EPRI TR-104788-R2, "PWR Primary-to-Secondary Leak Guidelines," Revision 2.6. NEI 97-06, "Steam Generator Program Guidelines." 7. UFSAR, Section 18, Table 18-1.8. Catawba License Renewal Commitments, CNS-1274.00-00-0016, Section 4.27.9. 10 CFR 50.67.Catawba Units 1 and 2 B 3.4.13-6 Revision No RCS PIV Leakage B 3.4.14 BASES ACTIONS (continued) | : 5. EPRI TR-104788-R2, "PWR Primary-to-Secondary Leak Guidelines," Revision 2.6. NEI 97-06, "Steam Generator Program Guidelines." 7. UFSAR, Section 18, Table 18-1.8. Catawba License Renewal Commitments, CNS-1274.00-00-0016, Section 4.27.9. 10 CFR 50.67.Catawba Units 1 and 2 B 3.4.13-6 Revision No RCS PIV Leakage B 3.4.14 BASES ACTIONS (continued) | ||
| Line 1,141: | Line 1,045: | ||
Accumulators B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued) | Accumulators B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued) | ||
SR 3.5.1.5 Verification e3a'that power is removed from each accumulator isolation valve operators for N154A, N165B, N176A, and N188B when the RCS pressure is > 1000 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, only two accumulators would be available for injection given a single failure coincident with a LOCA.'e -owe r is emoved a~l circui kers doc-ked lnder/control, t l~"31 da will, pro e adeg{uate a uranc hat powe " remove .-This SR allows power to be supplied to the motor operated isolation valves when RCS pressure is < 1000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. | SR 3.5.1.5 Verification e3a'that power is removed from each accumulator isolation valve operators for N154A, N165B, N176A, and N188B when the RCS pressure is > 1000 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, only two accumulators would be available for injection given a single failure coincident with a LOCA.'e -owe r is emoved a~l circui kers doc-ked lnder/control, t l~"31 da will, pro e adeg{uate a uranc hat powe " remove .-This SR allows power to be supplied to the motor operated isolation valves when RCS pressure is < 1000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. | ||
Even with power supplied to the'valves, inadvertent closure is prevented by the RCS pressure interlock associated with the valves.Should closure of a valve occur in spite of the interlock, the SI signal provided to the valves would open a closed valve in the event of a LOCA.REFERENCES | Even with power supplied to the'valves, inadvertent closure is prevented by the RCS pressure interlock associated with the valves.Should closure of a valve occur in spite of the interlock, the SI signal provided to the valves would open a closed valve in the event of a LOCA.REFERENCES | ||
: 1. IEEE Standard 279-1971 2. UFSAR, Chapter 6.3. 10 CFR 50.46.4. DPC-NE-3004. | : 1. IEEE Standard 279-1971 2. UFSAR, Chapter 6.3. 10 CFR 50.46.4. DPC-NE-3004. | ||
: 5. 10 CFR 50.36, Technical Specification, (c)(2)(ii). | : 5. 10 CFR 50.36, Technical Specification, (c)(2)(ii). | ||
| Line 1,170: | Line 1,074: | ||
Surveillance performance does not require removal of any tophat modules or grating, but the strainer exteriors shall be visually inspected. | Surveillance performance does not require removal of any tophat modules or grating, but the strainer exteriors shall be visually inspected. | ||
This surveillance is not a commitment to inspect 100% of the surface area of all tophats, but a sufficiently detailed inspection of exterior strainer surfaces is required to establish a high confidence that no adverse conditions are present. The scope of inspection necessary to provide high confidence includes 100% of the strainer areas that can be accessed and inspected using normal means and tools (i.e., flashlight, extendable mirror, hand held digital camera) without disassembly, and that difficult to access areas will be inspected to the extent possible using these same means.Any damage detected in the strainer assembly inspection will result in an expansion of the scope of the inspection to include other areas of potential damage. Inspection scope should be expanded, as needed, for degradation of strainer components identified during this inspection that were not considered readily accessible during the inspector's initial evaluation. | This surveillance is not a commitment to inspect 100% of the surface area of all tophats, but a sufficiently detailed inspection of exterior strainer surfaces is required to establish a high confidence that no adverse conditions are present. The scope of inspection necessary to provide high confidence includes 100% of the strainer areas that can be accessed and inspected using normal means and tools (i.e., flashlight, extendable mirror, hand held digital camera) without disassembly, and that difficult to access areas will be inspected to the extent possible using these same means.Any damage detected in the strainer assembly inspection will result in an expansion of the scope of the inspection to include other areas of potential damage. Inspection scope should be expanded, as needed, for degradation of strainer components identified during this inspection that were not considered readily accessible during the inspector's initial evaluation. | ||
REFERENCES | REFERENCES | ||
: 1. 10 CFR 50, Appendix A, GDC 35.2. 10 CFR 50.46.3. UFSAR, Section 6.2.1.4. UFSAR, Chapter 15.5. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. 10 CFR 50, Appendix A, GDC 35.2. 10 CFR 50.46.3. UFSAR, Section 6.2.1.4. UFSAR, Chapter 15.5. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
: 6. NRC Memorandum to V. Stello, Jr., from R.L. Baer,"Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.7. IE Information Notice No. 87-01.Catawba Units 1 and 2 B 3.5.2-10 Revision No I1 RWST B 3.5.4 BASES ACTIONS (continued) | : 6. NRC Memorandum to V. Stello, Jr., from R.L. Baer,"Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.7. IE Information Notice No. 87-01.Catawba Units 1 and 2 B 3.5.2-10 Revision No I1 RWST B 3.5.4 BASES ACTIONS (continued) | ||
| Line 1,242: | Line 1,146: | ||
-SR 3.6.8.2 Verifying HSS fan motor current at rated speed with the motor operated suction valves closed is indicative of overall fan motor performance. | -SR 3.6.8.2 Verifying HSS fan motor current at rated speed with the motor operated suction valves closed is indicative of overall fan motor performance. | ||
Since these fans are required to function during post-accident situations, the air density that the fans experience during surveillance testing will be different than the air density following a LOCA. An air density adjustment will be made to the average fan motor current test data before it is compared to the Technical Specification SR acceptance criteria. | Since these fans are required to function during post-accident situations, the air density that the fans experience during surveillance testing will be different than the air density following a LOCA. An air density adjustment will be made to the average fan motor current test data before it is compared to the Technical Specification SR acceptance criteria. | ||
Such inservice tests confirm component OPERABILITY, trend performance and detect incipient failures by indicating, abnormal performance. | Such inservice tests confirm component OPERABILITY, trend performance and detect incipient failures by indicating, abnormal performance. | ||
("h-)Frequency,4f 92 days w based on eratn expernce whic J showns Frequenc4o be accep.le SR 3.6.8.3 This SR verifies the motor operated suction valves open upon receipt of a Containment Pressure -High High signal and associated time delay and that the HSS fans receive a start permissive when the valves start to opn e ec-y ot was -base~ln opnc fig Caawb U 1 ands 2 uencyBto3be.68-ta4 Revision Catawba Units 1 and 2 B 3.6.8-4 Revision No.{,9) | ("h-)Frequency,4f 92 days w based on eratn expernce whic J showns Frequenc4o be accep.le SR 3.6.8.3 This SR verifies the motor operated suction valves open upon receipt of a Containment Pressure -High High signal and associated time delay and that the HSS fans receive a start permissive when the valves start to opn e ec-y ot was -base~ln opnc fig Caawb U 1 ands 2 uencyBto3be.68-ta4 Revision Catawba Units 1 and 2 B 3.6.8-4 Revision No.{,9) | ||
HSS B 3.6.8 BASES SURVEILLANCE REQUIREMENTS (continued) | HSS B 3.6.8 BASES SURVEILLANCE REQUIREMENTS (continued) | ||
| Line 1,257: | Line 1,161: | ||
The 1700°F temperature is a surveillance requirement. "An Analysis of Hydrogen Control Measures at McGuire Nuclear Station" (Ref. 5) section 3.8 identifies that the required normal operation temperature is 1500'F. Therefore, based upon ignitor performance testing conducted at Catawba, the surveillance requirement of 1700'F ensures that sufficient margin is resent for continued hydrogen ignition under degraded bus conditions.2. | The 1700°F temperature is a surveillance requirement. "An Analysis of Hydrogen Control Measures at McGuire Nuclear Station" (Ref. 5) section 3.8 identifies that the required normal operation temperature is 1500'F. Therefore, based upon ignitor performance testing conducted at Catawba, the surveillance requirement of 1700'F ensures that sufficient margin is resent for continued hydrogen ignition under degraded bus conditions.2. | ||
he 1 month Frequ is based on.the e1d tC orm 56 urveianca S der the ccnc s that apply(5.ing a Ant outagfand then potoial for an unpre ned transient if N furveill were/l erformed wi the reactor at .ower. OPerating,, exp~er" nce has own that t ise cornponen tus~ually pass the I-R when pe rmed at ine 18 mont.requenc, w is based fueling cCble. aherore, theUFrnitues w1 a 2 B 3o beviao le from a/alia bilitY 'a nd | he 1 month Frequ is based on.the e1d tC orm 56 urveianca S der the ccnc s that apply(5.ing a Ant outagfand then potoial for an unpre ned transient if N furveill were/l erformed wi the reactor at .ower. OPerating,, exp~er" nce has own that t ise cornponen tus~ually pass the I-R when pe rmed at ine 18 mont.requenc, w is based fueling cCble. aherore, theUFrnitues w1 a 2 B 3o beviao le from a/alia bilitY 'a nd | ||
~ REFERENCES | ~ REFERENCES | ||
: 1. 10 CF.R 50.44.2. 10 CFR 50, Appendix A, GDC 41.3. UFSAR, Section 6.2.4. 10 CFR 50-86, Technical Specifications, (c)(2)(ii). | : 1. 10 CF.R 50.44.2. 10 CFR 50, Appendix A, GDC 41.3. UFSAR, Section 6.2.4. 10 CFR 50-86, Technical Specifications, (c)(2)(ii). | ||
: 5. An Analysis of Hydrogen Control'Measures at.McGuire Nuclear Station-Catawba Units 1 and 2 B 3.6-9-5 R evisionNo AVS B 3.6.10 BASES ACTIONS (continued) | : 5. An Analysis of Hydrogen Control'Measures at.McGuire Nuclear Station-Catawba Units 1 and 2 B 3.6-9-5 R evisionNo AVS B 3.6.10 BASES ACTIONS (continued) | ||
| Line 1,288: | Line 1,192: | ||
Since these fans are required to function during post-accident situations, the air density that the fans experience during surveillance testing will be different than the air density following a LOCA. An air density adjustment will be made to the average fan motor current test data before it is compared to the Technical Specification SR acceptance criteria. | Since these fans are required to function during post-accident situations, the air density that the fans experience during surveillance testing will be different than the air density following a LOCA. An air density adjustment will be made to the average fan motor current test data before it is compared to the Technical Specification SR acceptance criteria. | ||
Such inservice tests confirm component OPERABILITY, .trend performace | Such inservice tests confirm component OPERABILITY, .trend performace | ||
_an frt incipient failures by indicating abnormal performane. | _an frt incipient failures by indicating abnormal performane. | ||
{The Frequ. e of-f92 days c rms with the reurmet' sýýlrES ýc Sequip :nt and considersAe known re~abaaili, yof fan controls, a he two train red,, dancy available INS ,RT , SR 3.6.11.3 Verifying the OPERABILITY of the return air damper provides assurance that the proper flow path will exist when the fan is started. This Surveillance also tests the circuitry, including time delays to ensure the sytem operates properly. | {The Frequ. e of-f92 days c rms with the reurmet' sýýlrES ýc Sequip :nt and considersAe known re~abaaili, yof fan controls, a he two train red,, dancy available INS ,RT , SR 3.6.11.3 Verifying the OPERABILITY of the return air damper provides assurance that the proper flow path will exist when the fan is started. This Surveillance also tests the circuitry, including time delays to ensure the sytem operates properly. | ||
[T Frequen of 92 days d ev-elofp~e~d.,/ | [T Frequen of 92 days d ev-elofp~e~d.,/ | ||
co0nsi ring t importan/ | co0nsi ring t importan/ | ||
| Line 1,325: | Line 1,229: | ||
Additionally, the minimum boron concentration setpoint is used to assure reactor subcriticality in a post LOCA environment, while the maximum boron concentration is used as the bounding value in the hot leg switchover timing calculation (Ref.4). This is accomplished by obtaining at least 24 ice samples. Each sample is taken approximately one foot from. the top of the ice of each randomly selected ice basket in each ice condenser bay. The SR is modified by a NOTE that allows the boron concentration and pH value obtained from averaging the individual samples' analysis results to satisfy the requirements of the SR. If either the average boron concentration or average pH value is outside their prescribed limit, then entry into ACTION Condition A is required. | Additionally, the minimum boron concentration setpoint is used to assure reactor subcriticality in a post LOCA environment, while the maximum boron concentration is used as the bounding value in the hot leg switchover timing calculation (Ref.4). This is accomplished by obtaining at least 24 ice samples. Each sample is taken approximately one foot from. the top of the ice of each randomly selected ice basket in each ice condenser bay. The SR is modified by a NOTE that allows the boron concentration and pH value obtained from averaging the individual samples' analysis results to satisfy the requirements of the SR. If either the average boron concentration or average pH value is outside their prescribed limit, then entry into ACTION Condition A is required. | ||
Sodium tetraborate has been proven effective in maintaining the boron content for long storage periods, and it also enhances the ability of the solution to remove and retain fission product iodine. The high pH is required to enhance the effectiveness of the ice-- ,- and the melted ice in removing iodine froý i containment-atmosphere..- | Sodium tetraborate has been proven effective in maintaining the boron content for long storage periods, and it also enhances the ability of the solution to remove and retain fission product iodine. The high pH is required to enhance the effectiveness of the ice-- ,- and the melted ice in removing iodine froý i containment-atmosphere..- | ||
This pH range also minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to ECCS and Containment S ray Sstem fluids in the recirculation mode of op:eration. | This pH range also minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to ECCS and Containment S ray Sstem fluids in the recirculation mode of op:eration. | ||
[The Frequency of 54 mo ths is intendedt withexpected le 'gth of three fuel c les, and was devel-ed cnsidering these facts: a. Lon/term ice storage t dsts have, determine hat the chemical /cc,onposition of the sted ice is extremelyable; | [The Frequency of 54 mo ths is intendedt withexpected le 'gth of three fuel c les, and was devel-ed cnsidering these facts: a. Lon/term ice storage t dsts have, determine hat the chemical /cc,onposition of the sted ice is extremelyable; | ||
' /b. *'here are no nor (al operating, mech i~sms that significar y -change the bor concentration of t stored ice, and remains within a 9.0 -.5 range when bo concentrations e above approximat 1200 ppm; and Catawba Units 1 and 2 B 3.6.12-10 Revision No.f Ice Bed B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued) nce has d nstrated that m the boro pH r irements has no ,en a problem]REFERENCES | ' /b. *'here are no nor (al operating, mech i~sms that significar y -change the bor concentration of t stored ice, and remains within a 9.0 -.5 range when bo concentrations e above approximat 1200 ppm; and Catawba Units 1 and 2 B 3.6.12-10 Revision No.f Ice Bed B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued) nce has d nstrated that m the boro pH r irements has no ,en a problem]REFERENCES | ||
: 1. UFSAR, Section 6.2.2. 10 CFR 50, Appendix K.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. UFSAR, Section 6.2.2. 10 CFR 50, Appendix K.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
: 4. UFSAR, Section 6.3.3.5. Topical Report ICUG-001, Application of the Active Ice Mass Management Concept to the Ice Condenser Ice Mass Technical Specification, Revision 2.6. UFSAR, Section 18, Table 18-1.7. Catawba License Renewal Commitments, CNS-1274.00-00-0016, Section 4.17.Catawba Units 1 and 2 B 3.6.12-11 Revision No. 0 Ice Condenser Doors B 3.6.13 BASES ACTIONS (continued) 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE REQUIREMENTS SR 3.6.13.1 Verifying, by means of the Inlet Door Position Monitoring System, that the inlet doors are in their closed positions makes the operator aware of an inadvertent opening of one or more doors. JT-6 Freque~n t.,2hour_ensuresýare of thý,e eaust a e twh erd oor +k SR 3.6.13.2 Verifying, by visual inspection, that each intermediate deck door is closed and not impaired by ice, frost, or debris provides assurance that the intermediate deck doors (which form the floor of the upper plenum where frequent maintenance on the ice bed is performed) have not been left open or obstructed. | : 4. UFSAR, Section 6.3.3.5. Topical Report ICUG-001, Application of the Active Ice Mass Management Concept to the Ice Condenser Ice Mass Technical Specification, Revision 2.6. UFSAR, Section 18, Table 18-1.7. Catawba License Renewal Commitments, CNS-1274.00-00-0016, Section 4.17.Catawba Units 1 and 2 B 3.6.12-11 Revision No. 0 Ice Condenser Doors B 3.6.13 BASES ACTIONS (continued) 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE REQUIREMENTS SR 3.6.13.1 Verifying, by means of the Inlet Door Position Monitoring System, that the inlet doors are in their closed positions makes the operator aware of an inadvertent opening of one or more doors. JT-6 Freque~n t.,2hour_ensuresýare of thý,e eaust a e twh erd oor +k SR 3.6.13.2 Verifying, by visual inspection, that each intermediate deck door is closed and not impaired by ice, frost, or debris provides assurance that the intermediate deck doors (which form the floor of the upper plenum where frequent maintenance on the ice bed is performed) have not been left open or obstructed. | ||
| Line 1,357: | Line 1,261: | ||
: 1. UFSAR, Section 6.2.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. UFSAR, Section 6.2.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3.6.14-6 Revision No.10 Containment Recirculation Drains BASES B 3.6.15 SURVEILLANCE SR 3.6.15.1 and SR 3.6.15.2 REQUIREMENTS Verifying the OPERABILITY of the refueling canal drains ensures that they will be able to perform their functions in the event of a DBA. SR 3.6.15.1 confirms that the refueling canal drain valves have been locked open and that the drains are clear of any obstructions that could impair their functioning. | Catawba Units 1 and 2 B 3.6.14-6 Revision No.10 Containment Recirculation Drains BASES B 3.6.15 SURVEILLANCE SR 3.6.15.1 and SR 3.6.15.2 REQUIREMENTS Verifying the OPERABILITY of the refueling canal drains ensures that they will be able to perform their functions in the event of a DBA. SR 3.6.15.1 confirms that the refueling canal drain valves have been locked open and that the drains are clear of any obstructions that could impair their functioning. | ||
In addition to debris near the drains, SR 3.6.15.2 requires attention be given to any debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation and water is flowing to the drains. SR 3.6.15.1 must be performed before entering MODE 4 from MODE 5 after every filling of the canal to ensure that the valves have been locked open and that no debris that cou, impairthe drainsawas deposited durinn the time the canal was filled. SR 3.6A5.2 is performed ev 92 daysforthe uooDeq om artme t and refuel canal a as. 'The 92 Frequency s develop considering such ctors as the ccessibility of e drain Lthe ab f4'ratf. in M vicinity of tho06rains and th edunda of ZE J12 T2Z-SR 3.6.15.3 Verifying the OPERABILITY of the ice condenser floor drains ensures that they will be able to perform their functions in the event of a DBA.Inspecting the drain valve disk ensures that the valve is performing its function of sealing the drain line from warm air leakage into the ice condenser during normal operation, yet will open if melted ice fills the line following a DBA. Verifying that the drain lines are not obstructed ensures their, readiness to drain water from the ice cond~ennser.[-The 18 lFre develop, considerin-, such factors as'the inaccessi lity of tl dtrains durina ower operati n; the design of~the ice con riser, ,[ whn th'Sureilanceis/prfored t f_ 18monhe Freq Beas of high radiation in the vicinity of the drains during power oper tion, this Surveillance is normally done during a shutdown.REFERENCES | In addition to debris near the drains, SR 3.6.15.2 requires attention be given to any debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation and water is flowing to the drains. SR 3.6.15.1 must be performed before entering MODE 4 from MODE 5 after every filling of the canal to ensure that the valves have been locked open and that no debris that cou, impairthe drainsawas deposited durinn the time the canal was filled. SR 3.6A5.2 is performed ev 92 daysforthe uooDeq om artme t and refuel canal a as. 'The 92 Frequency s develop considering such ctors as the ccessibility of e drain Lthe ab f4'ratf. in M vicinity of tho06rains and th edunda of ZE J12 T2Z-SR 3.6.15.3 Verifying the OPERABILITY of the ice condenser floor drains ensures that they will be able to perform their functions in the event of a DBA.Inspecting the drain valve disk ensures that the valve is performing its function of sealing the drain line from warm air leakage into the ice condenser during normal operation, yet will open if melted ice fills the line following a DBA. Verifying that the drain lines are not obstructed ensures their, readiness to drain water from the ice cond~ennser.[-The 18 lFre develop, considerin-, such factors as'the inaccessi lity of tl dtrains durina ower operati n; the design of~the ice con riser, ,[ whn th'Sureilanceis/prfored t f_ 18monhe Freq Beas of high radiation in the vicinity of the drains during power oper tion, this Surveillance is normally done during a shutdown.REFERENCES | ||
: 1. UFSAR, Section 6.2.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. UFSAR, Section 6.2.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3- 15-4 Revision No.0 Reactor Building B 3.6.16 BASES APPLICABILITY Maintaining reactor building OPERABILITY prevents leakage of radioactive material from the reactor building. | Catawba Units 1 and 2 B 3- 15-4 Revision No.0 Reactor Building B 3.6.16 BASES APPLICABILITY Maintaining reactor building OPERABILITY prevents leakage of radioactive material from the reactor building. | ||
| Line 1,373: | Line 1,277: | ||
0 eraing epeece hs- s wn--'thes componen~isUsually pa 'te Surveilla ce when perfo led at}the 1!'onthr n The Fquen~cy is a eptable from feliability/ | 0 eraing epeece hs- s wn--'thes componen~isUsually pa 'te Surveilla ce when perfo led at}the 1!'onthr n The Fquen~cy is a eptable from feliability/ | ||
SR 3.7.4.3 The function of the block valve is.to isolate a failed open SG PORV.Cycling the block valve both closed and open demonstrates its capability to perform this function. | SR 3.7.4.3 The function of the block valve is.to isolate a failed open SG PORV.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-unit cooldown may satisfy this requiremet pte -_-g e --ienceh s own-that lse compon ms'usualy t-he ---IqSurveillai perf 'ded at the 1 "onth Frequ h LFrequ cy is accept e from a rel*ility standlt.eft REFERENCES | Performance of inservice testing or use of the block valve during-unit cooldown may satisfy this requiremet pte -_-g e --ienceh s own-that lse compon ms'usualy t-he ---IqSurveillai perf 'ded at the 1 "onth Frequ h LFrequ cy is accept e from a rel*ility standlt.eft REFERENCES | ||
: 1. UFSAR, Section 10.3.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. UFSAR, Section 10.3.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3.7.4-4 Revision Noe AFW System B 3.7.5 BASES ACTIONS (continued) | Catawba Units 1 and 2 B 3.7.4-4 Revision Noe AFW System B 3.7.5 BASES ACTIONS (continued) | ||
| Line 1,429: | Line 1,333: | ||
ani o ssýwith the typical SR 3.7.10.4 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 consequences is no more than 5 rem TEDE and the CRE occupants are protected from hazardous chemicals and smoke. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences. | ani o ssýwith the typical SR 3.7.10.4 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 consequences is no more than 5 rem TEDE and the CRE occupants are protected from hazardous chemicals and smoke. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences. | ||
When unfiltered air inleakage is greater than the assumed flow rate, Condition B Catawba Units 1 and 2 B 3.7:10-7 Revision CRAVS: = T2S PA .I B 3-7-10 BASES SURVEILLANCE REQUIREMENTS (continued) must be entered- Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for the occupants following an accident. | When unfiltered air inleakage is greater than the assumed flow rate, Condition B Catawba Units 1 and 2 B 3.7:10-7 Revision CRAVS: = T2S PA .I B 3-7-10 BASES SURVEILLANCE REQUIREMENTS (continued) must be entered- Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for the occupants following an accident. | ||
Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3 (Ref- 9), which endorses, with exceptions, NEI 99-03, Section 8-4 and Appendix F (Ref. 7). These compensatory measures may also be used as mitigating actions as required by Required Action B.2. Temporary analytical methods may also be used as compensatory measures to restore OPERABILITY (Ref.8). Options for restoring the CRE boundary to OPERABLE status include changing the licensing basis DBA consequence analysis, repairing the CRE boundary, or a combination of these actions. Depending upon the nature of the problem and the corrective action, a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status.REFERENCES | Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3 (Ref- 9), which endorses, with exceptions, NEI 99-03, Section 8-4 and Appendix F (Ref. 7). These compensatory measures may also be used as mitigating actions as required by Required Action B.2. Temporary analytical methods may also be used as compensatory measures to restore OPERABILITY (Ref.8). Options for restoring the CRE boundary to OPERABLE status include changing the licensing basis DBA consequence analysis, repairing the CRE boundary, or a combination of these actions. Depending upon the nature of the problem and the corrective action, a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status.REFERENCES | ||
: 1. UFSAR, Section 6.4.2. UFSAR, Section 9.4.1.3. .UFSAR, Chapter 15.4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. UFSAR, Section 6.4.2. UFSAR, Section 9.4.1.3. .UFSAR, Chapter 15.4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
: 5. Regulatory Guide 1.52, Rev. 2.6. Catawba Nuclear Station License Amendments 90/84 for Units 1/2, August 23, 1991.7. NEI 99-03, "Control Room Habitability Assessment", June 2001.8. Letter from Eric J. Leeds (NRC) to James W. Davis (NEI) dated January 30, 2004, "NEI Draft White Paper, Use of Generic Letter 91-18 Process and Alternative Source Terms in the Context of Control Room Habitability", (ADAMS Accession No.ML040300694). | : 5. Regulatory Guide 1.52, Rev. 2.6. Catawba Nuclear Station License Amendments 90/84 for Units 1/2, August 23, 1991.7. NEI 99-03, "Control Room Habitability Assessment", June 2001.8. Letter from Eric J. Leeds (NRC) to James W. Davis (NEI) dated January 30, 2004, "NEI Draft White Paper, Use of Generic Letter 91-18 Process and Alternative Source Terms in the Context of Control Room Habitability", (ADAMS Accession No.ML040300694). | ||
: 9. Regulatory Guide 1.196, Rev. 1.Catawba Units 1 and 2 B 3.7.10-8 Revision No. 0 CRACWS B 3.7.11 BASES SURVEILLANCE REQUIREMENTS SR 3.7.11.1 This SR verifies that the heat removal capability of the system is s'to maintain the temperature in the control room.a or below 0F/2r-re-quejp'y is appropri'e since signifi nt degrada5n of-theslow and i.ot expectedcrý r this time p"riod.-REFERENCES | : 9. Regulatory Guide 1.196, Rev. 1.Catawba Units 1 and 2 B 3.7.10-8 Revision No. 0 CRACWS B 3.7.11 BASES SURVEILLANCE REQUIREMENTS SR 3.7.11.1 This SR verifies that the heat removal capability of the system is s'to maintain the temperature in the control room.a or below 0F/2r-re-quejp'y is appropri'e since signifi nt degrada5n of-theslow and i.ot expectedcrý r this time p"riod.-REFERENCES | ||
: 1. UFSAR, Section 9.4.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. UFSAR, Section 9.4.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
: 3. 10 CFR 50.67, Accident source term.4. Regulatory Guide 1.183, Revision 0.Catawba Units 1 and 2 B 3.7.11-4 Revision No 19 4) | : 3. 10 CFR 50.67, Accident source term.4. Regulatory Guide 1.183, Revision 0.Catawba Units 1 and 2 B 3.7.11-4 Revision No 19 4) | ||
| Line 1,443: | Line 1,347: | ||
As the environment and normal operating conditions on this system are not severe, testing each train once a month provides an adequate check on this system. Monthly heater operations dry out any moisture that may have accumulated in the carbon from humidity in the ambient air. Systems with heaters must be operated from the control room: _ 10 continuous hours with flow through the HEPA filters and Catawba Units 1 and 2 B 3.7.12-4 Revision No. 2 ABFVES B 3.7.12 BASES SURVEILLANCE REQUIREMENTS (continued) cabnasres and with the he-atrs energized. | As the environment and normal operating conditions on this system are not severe, testing each train once a month provides an adequate check on this system. Monthly heater operations dry out any moisture that may have accumulated in the carbon from humidity in the ambient air. Systems with heaters must be operated from the control room: _ 10 continuous hours with flow through the HEPA filters and Catawba Units 1 and 2 B 3.7.12-4 Revision No. 2 ABFVES B 3.7.12 BASES SURVEILLANCE REQUIREMENTS (continued) cabnasres and with the he-atrs energized. | ||
The 31 y Frequency? | The 31 y Frequency? | ||
SR 3.7.12.2 This SR verifies that the required ABFVES testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The ABFVES filter tests are in accordance with Reference | SR 3.7.12.2 This SR verifies that the required ABFVES testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The ABFVES filter tests are in accordance with Reference | ||
: 5. The VFTP includes testing HEPA filter performance, carbon adsorbers efficiency, system flow rate, and the physical properties of the activated carbon (general use and following specific operations). | : 5. The VFTP includes testing HEPA filter performance, carbon adsorbers efficiency, system flow rate, and the physical properties of the activated carbon (general use and following specific operations). | ||
The system flow rate determination and in-place testing of the filter unit components is performed in the normal operating alignment with both trains in operation. | The system flow rate determination and in-place testing of the filter unit components is performed in the normal operating alignment with both trains in operation. | ||
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Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies sufficient spent fuel pool water is available in the event of a fuel handling accident. | Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies sufficient spent fuel pool water is available in the event of a fuel handling accident. | ||
The water level in the spent fuel pool must be The 74 Frequency is ap riate because t f-olume pool isnormy stable. Water,/el changes are c trolled byUplarIjafrocedures a are acceptable ed oe During refueling operations, the level in the spent fuel pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.Catawba Units 1 and 2 B 3.7.14-2 -Revision Not Spent Fuel Pool Boron Concentration B 3.7.15 BASES APPLICABLE SAFETY ANALYSES (continued) | The water level in the spent fuel pool must be The 74 Frequency is ap riate because t f-olume pool isnormy stable. Water,/el changes are c trolled byUplarIjafrocedures a are acceptable ed oe During refueling operations, the level in the spent fuel pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.Catawba Units 1 and 2 B 3.7.14-2 -Revision Not Spent Fuel Pool Boron Concentration B 3.7.15 BASES APPLICABLE SAFETY ANALYSES (continued) | ||
The concentration of dissolved boron in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 7).LCO The spent fuel pool boron concentration is required to be within the limits specified in the COLR. The specified concentration of dissolved boron in the spent fuel pool preserves the assumptions used in the analyses of the potential critical accident scenarios as described in Reference | The concentration of dissolved boron in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 7).LCO The spent fuel pool boron concentration is required to be within the limits specified in the COLR. The specified concentration of dissolved boron in the spent fuel pool preserves the assumptions used in the analyses of the potential critical accident scenarios as described in Reference | ||
: 6. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the spent fuel pool.APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.When the concentration of boron in the fuel storage pool 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. | : 6. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the spent fuel pool.APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.When the concentration of boron in the fuel storage pool 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 concentration of boron is restored simultaneously with suspending movement of fuel assemblies. | The concentration of boron is restored simultaneously with suspending movement of fuel assemblies. | ||
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LCO 3.0.3 must be entered immediately to commence a controlled shutdown.SURVEILLANCE SR 3.8.9.1 REQUIREMENTS This Surveillance verifies that the AC, channels of DC, DC trains, and AC vital bus electrical power distribution systems are functioning properly, with the correct circuit breaker alignment. | LCO 3.0.3 must be entered immediately to commence a controlled shutdown.SURVEILLANCE SR 3.8.9.1 REQUIREMENTS This Surveillance verifies that the AC, channels of DC, DC trains, and AC vital bus electrical power distribution systems are functioning properly, with the correct circuit breaker alignment. | ||
The correct breaker alignment ensures the appropriate separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required bus. The verification of proper indicated voltage availability on the buses ensures that the required voltage is readily available for motive as well as control functions for critical system loads connected to these bue.Te 7 daFrequency takes int pccodunffte-o/-f the AC,D, and AC vital bus e rical power distribýW ubsys;e s, and ot indications availa tirQmthatfalert theoperator/ | The correct breaker alignment ensures the appropriate separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required bus. The verification of proper indicated voltage availability on the buses ensures that the required voltage is readily available for motive as well as control functions for critical system loads connected to these bue.Te 7 daFrequency takes int pccodunffte-o/-f the AC,D, and AC vital bus e rical power distribýW ubsys;e s, and ot indications availa tirQmthatfalert theoperator/ | ||
to sytem REFERENCES | to sytem REFERENCES | ||
: 1. UFSAR, Chapter 6.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. UFSAR, Chapter 6.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
: 4. Regulatory Guide-1.93, December 1974.Catawba Units 1 and 2 B 3.8.9-9 Revision Distribution Systems-Shutdown B 3.8.10 BASES SURVEILLANCE SR 3.8.10.1 REQUIREMENTS This Surveillance verifies that the AC, channels of DC, DC trains, and AC vital bus electrical power distribution subsystems are functioning properly, with all the buses energized. | : 4. Regulatory Guide-1.93, December 1974.Catawba Units 1 and 2 B 3.8.9-9 Revision Distribution Systems-Shutdown B 3.8.10 BASES SURVEILLANCE SR 3.8.10.1 REQUIREMENTS This Surveillance verifies that the AC, channels of DC, DC trains, and AC vital bus electrical power distribution subsystems are functioning properly, with all the buses energized. | ||
The verification of proper indicated 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. 7 day Frequen taes into a.p.oeount the-ýiaiýý fte ectnca ;zwer distributio 5Obsystems, other v (indicatio available in e. ontrol roor vtat aler the Zlerator to/subs ~em malfunc *"ns. i-REFERENCES | The verification of proper indicated 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. 7 day Frequen taes into a.p.oeount the-ýiaiýý fte ectnca ;zwer distributio 5Obsystems, other v (indicatio available in e. ontrol roor vtat aler the Zlerator to/subs ~em malfunc *"ns. i-REFERENCES | ||
: 1. UFSAR, Chapter 6.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. UFSAR, Chapter 6.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
Catawba Units 1 and 2 B 3.8.10-4 Revision NO CHANGES THIS PACE.O INFORMATiOw ONLY Boron Concentration B 3.9.1 BASES ACTIONS A.1 and A.2 Continuation of CORE ALTERATIONS or positive reactivity additions (including actions to reduce boron concentration) is contingent upon maintaining the unit in compliance with the LCO. If the boron concentration of any coolant volume in the RCS, the refueling canal, or the refueling cavity is less than its limit, all operations involving CORE ALTERATIONS or positive reactivity additions must be suspended immediately. | Catawba Units 1 and 2 B 3.8.10-4 Revision NO CHANGES THIS PACE.O INFORMATiOw ONLY Boron Concentration B 3.9.1 BASES ACTIONS A.1 and A.2 Continuation of CORE ALTERATIONS or positive reactivity additions (including actions to reduce boron concentration) is contingent upon maintaining the unit in compliance with the LCO. If the boron concentration of any coolant volume in the RCS, the refueling canal, or the refueling cavity is less than its limit, all operations involving CORE ALTERATIONS or positive reactivity additions must be suspended immediately. | ||
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The Surva 2e interval is/ ,811ected to be co surate with the normal duratio .rtime to complete/ uel handlinog. | The Surva 2e interval is/ ,811ected to be co surate with the normal duratio .rtime to complete/ uel handlinog. | ||
erations. | erations. | ||
As such, this Surveilla oe. ensures that a "_ postulate tfuel handling accident involving recently irradiated fuel thatJ Catawba Units 1 and 2 B 3.9.3-4 Revision Containment Penetrations B 3.9.3 BASES SURVEILLANCE REQUIREMENTS (continued) release$ssion product dioactivity.withj thecont ent will t result inea rease of signific t fission prod radioaOyit to the e ironme SR 3.9.3.2 Standby systems should be checked periodically to ensure that they function properly s nvi nment an nto _.r.g conditi0ns on}no] sev eesting each trai e month-provL~ies ansa ThiseSRaverifiecs th0t is system. M eatperforatimei ry out any wt SoiVre that m Fy have a ccumulaed in the carbon from humidity in the Exhaus .Systems with heaters n ardust ne operated by initifatinc through the HEPA filters and activated carbon adsorbers for > 1 continuous hours with the heaters of. tay Fre Yasc carso hene n re anflin ispmenitf n e two traedundanci t es SR 3.9.3.3 This SR verifies that the required testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The Containment Purge Exhaust System filter tests are in accordance with Reference | As such, this Surveilla oe. ensures that a "_ postulate tfuel handling accident involving recently irradiated fuel thatJ Catawba Units 1 and 2 B 3.9.3-4 Revision Containment Penetrations B 3.9.3 BASES SURVEILLANCE REQUIREMENTS (continued) release$ssion product dioactivity.withj thecont ent will t result inea rease of signific t fission prod radioaOyit to the e ironme SR 3.9.3.2 Standby systems should be checked periodically to ensure that they function properly s nvi nment an nto _.r.g conditi0ns on}no] sev eesting each trai e month-provL~ies ansa ThiseSRaverifiecs th0t is system. M eatperforatimei ry out any wt SoiVre that m Fy have a ccumulaed in the carbon from humidity in the Exhaus .Systems with heaters n ardust ne operated by initifatinc through the HEPA filters and activated carbon adsorbers for > 1 continuous hours with the heaters of. tay Fre Yasc carso hene n re anflin ispmenitf n e two traedundanci t es SR 3.9.3.3 This SR verifies that the required testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The Containment Purge Exhaust System filter tests are in accordance with Reference | ||
: 4. The VFTP includes testing HEPA filter performance, carbon adsorbers efficiency, system flow rate, and the physical properties of the activated carbon (general use and following specific operations). | : 4. The VFTP includes testing HEPA filter performance, carbon adsorbers efficiency, system flow rate, and the physical properties of the activated carbon (general use and following specific operations). | ||
Specific test Frequencies and additional information are discussed in detail in the VFTP.REFERENCES | Specific test Frequencies and additional information are discussed in detail in the VFTP.REFERENCES | ||
: 1. UFSAR, Section 15.7.4.2. NUREG-0800, Section 15.7.4, Rev. 1, July 1981.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | : 1. UFSAR, Section 15.7.4.2. NUREG-0800, Section 15.7.4, Rev. 1, July 1981.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii). | ||
: 4. Regulatory Guide 1.52 (Rev. 2).5. 10 CFR 50.67, Accident source term.6. Regulatory Guide 1.183 (Rev. 0).7. Catawba Nuclear Station License Amendments 90/84 for Units 1/2, August 23, 1991.Catawba Units 1 and 2 B 3.9.3-5 Revision Noj' RHR and Coolant Circulation-High Water Level B 3.9.4 BASES ACTIONS (continued) | : 4. Regulatory Guide 1.52 (Rev. 2).5. 10 CFR 50.67, Accident source term.6. Regulatory Guide 1.183 (Rev. 0).7. Catawba Nuclear Station License Amendments 90/84 for Units 1/2, August 23, 1991.Catawba Units 1 and 2 B 3.9.3-5 Revision Noj' RHR and Coolant Circulation-High Water Level B 3.9.4 BASES ACTIONS (continued) | ||
| Line 1,663: | Line 1,567: | ||
Closing containment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded.The Completion Time of 4 hours is reasonable, based on the low probability of the coolant boiling in that time.SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This Surveillance demonstrates that the RHR loop is in operation and circulating, reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. The RCS temperature is detrmined to an is aitaied. | Closing containment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded.The Completion Time of 4 hours is reasonable, based on the low probability of the coolant boiling in that time.SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This Surveillance demonstrates that the RHR loop is in operation and circulating, reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. The RCS temperature is detrmined to an is aitaied. | ||
of oTHurs is sufficiepýcnsie.o e-fow, tempe t_ rb, pump con~l.e and alarmn indi ons availa he....e controlro. | of oTHurs is sufficiepýcnsie.o e-fow, tempe t_ rb, pump con~l.e and alarmn indi ons availa he....e controlro. | ||
for monitoring t telHR -S Sstem- _ " --REFERENCES | for monitoring t telHR -S Sstem- _ " --REFERENCES | ||
: 1. UFSAR, Section 5.5.7.2. 10 CFR 50.36, Technical Specifications, (cX2)(ii). | : 1. UFSAR, Section 5.5.7.2. 10 CFR 50.36, Technical Specifications, (cX2)(ii). | ||
: 3. NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management." Catawba Units 1 and 2 B 3.9.4-4 Revision No RHR and Coolant Circulation-Low Water Level B 3.9.5 BASES ACTIONS (continued) concentration greater than that which would be required in the RCS for minimum refueling boron concentration. | : 3. NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management." Catawba Units 1 and 2 B 3.9.4-4 Revision No RHR and Coolant Circulation-Low Water Level B 3.9.5 BASES ACTIONS (continued) concentration greater than that which would be required in the RCS for minimum refueling boron concentration. | ||
Revision as of 04:52, 1 May 2019
| ML100920160 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/31/2010 |
| From: | Morris J R Duke Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML100920160 (348) | |
Text
DukeAMEs R MORRS Energy Vice President Duke Energy Corporation Catawba Nuclear Station 4800 Concord Road March 31, 2010 York, SC 29745 803-701-4251 803-701-3221 fax U. S. Nuclear Regulatory Commission ATTENTION:
Document Control Desk Washington, D.C. 20555 1 OCFR50.90
Subject:
Duke Energy Carolinas, LLC (Duke Energy)Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program In accordance with the provisions of 10 CFR 50.90, Duke Energy is submitting a request for an amendment to the Technical Specifications (TS) for Catawba Nuclear Station (Catawba)
Units 1 and 2.The proposed amendment would modify Catawba's Technical Specifications by relocating specific surveillance frequencies to a licensee controlled program with the implementation of Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specification Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies." The changes are consistent with the NRC approved Industry/TSTF Standard Technical Specification (STS) change TSTF-425, Revision 3 (ADAMS-Accession No. ML080280275).
Availability of this TSTF was published in the Federal Register notice on July 6, 2009.Attachment 1 provides a description of the proposed changes, the requested confirmation of applicability, and plant specific verifications.
Attachment 2 provides the Probabilistic Risk Assessment (PRA) technical adequacy documentation.
Attachment 3 provides:the existing TS pages marked up to illustrate the proposed changes. Attachment 4 provides the proposed TS Bases pages. Attachment 5 provides a cross reference table comparing the TSTF surveillance numbers to the Catawba surveillance numbers. Attachment 6 provides the Proposed No Significant Hazards Consideration.
Duke Energy requests the NRC's review and approval of the proposed license amendment within one ye'ar of this submittal.
Duke Energy is requesting a 90-day implementation grace period due to the extensive document changes necessary to implement this license amendment.
Also, Duke Energy will update applicable sections of the Catawba UFSAR, as necessary, and submit these changes per 1OCFR 50.71(e).In accordance with Duke Energy administrative procedures and the Quality Assurance Program Topical Report, this proposed amendment has been reviewed and approved by the Catawba Plant Operations Review Committee.
In accordance with 10 CFR 50.91, a copy of this proposed amendment is being forwarded to the appropriate South Carolina State officials. , 0)www. duke-energy comr March 31, 2010 Nuclear Regulatory Commission Page 2 There are no new commitments being made as a result of this proposed change.Inquiries regarding this submittal should be directed to Tony Jackson at 803-701-3742.
Sincerely, J aws R. Morris Attachments:
- 1. Description and Assessment
- 2. Documentation of PRA Technical Adequacy 3. Proposed Technical Specification Changes 4. Proposed Technical Specification Bases Changes 5. Surveillance Frequency Cross Reference Table 6. Proposed No Significant Hazards Consideration March 31, 2010 Nuclear Regulatory Commission Page 3 cc: w/Attachments L. A. -Reyes NRC Region II Administrator U.S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 J. H. Thompson Project Manager (Catawba)U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8-G9A 11555 Rockville Pike Rockville, MD 20852-2738 G. A. Hutto NRC Senior Resident Inspector Catawba Nuclear Station Susan E. Jenkins, Manager Radioactive
& Infectious Waste Management Division of Waste Management South Carolina Department of Health and Environmental Control 2600 Bull St.Columbia, SC 29201 March 31, 2010 Nuclear Regulatory Commission Page 4 OATH AND AFFIRMATION James R. Morris affirms that he is the person who subscribed his name to-the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.
Jame s, Site Vice President Subscribed and sworn to me: Notary Public Date D te hw~My commission expires:
ATTACHMENT 1 DESCRIPTION AND ASSESSMENT Catawba Nuclear Station Attachment 1 Description and Assessment 1.0 Description The proposed amendment would modify the Catawba Nuclear Station (Catawba)Technical Specifications (TS) by relocating specific 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." Additionally, the change would add a new program, the Surveillance Frequency Control Program, to TS Section 5.0, "Administrative Controls." The changes are consistent with theNRC approved Industry/TSTF Standard Technical Specification (STS) change TSTF-425, Revision 3 (ADAMS Accession No.ML080280275).
Availability of this TSTF was published in the Federal Register notice on July 6, 2009.2.0 Assessment 2.1 Applicability of Published Safety Evaluation Duke has reviewed the Safety Evaluation Report (SER) dated July 6, 2009. This included a review of the NRC staff's evaluation of TSTF-425, Revision 3, and the requirements specified in NEI 04-10, Rev. 1, (ADAMS Accession No. ML071360456).
Attachment 2 includes Duke Energy's documentation with regards to the PRA technical adequacy, consistent with the requirements of Regulatory Guide 1.200, Rev. 1 (ADAMS Accession No. ML070240001), Section 4.2, and describes any PRA models without NRC endorsed standards, including documentation of the quality characteristics of the models.Duke Energy has concluded that the justifications presented in the TSTF-425 proposal and the safety evaluation prepared by the NRC staff is applicable to Catawba Units 1 and 2, and justify this amendment to incorporate the changes to the Catawba TS.2.2 Optional Changes and Variations The proposed amendment is consistent with the STS changes described in TSTF-425, Revision 3, however, Duke Energy proposes variations or deviations from TSTF-425, as identified below.1. The revised (re-typed)
TS pages are not included in this proposed amendment due to the number of TS pages affected, the nature of the proposed changes, and the outstanding amendment requests that Catawba currently has under NRC review. Providing only the mark-ups of the proposed TS.changes satisfies the requirements of 10 CFR 50.90. This is an administrative deviation from TSTF-425 with no exceptions to the NRC staff's model safety evaluation dated July 6, 2009. This administrative deviation is consistent with Exelon's Peach Bottom Atomic Power Station License Amendment application dated August 31, 2009 (NRC Accession No. ML092470153).
Page 1 of 3 Catawba Nuclear Station 'Attachment 1 Description and Assessment
- 2. A note in Technical Specification 3.3.1 concerning the one-time extension for SR 3.3.1.5 will be deleted since it has expired and the same page is being revised for this amendment.
- 3. Attachment 5 provides a cross reference table between the NUREG-1431 surveillances included in TSTF-425 versus the Catawba surveillances included in this amendment request. This cross reference table highlights the following:
- a. TSTF-425 surveillances with identical corresponding Catawba Surveillance numbers, b. TSTF-425 surveillances and corresponding Catawba Surveillances but with differing Surveillance numbers, c. TSTF-425 surveillances that are not contained in the Catawba TS and therefore not applicable, and d. Catawba plant specific surveillances that are not contained in TSTF-425 surveillance mark-ups, but are applicable to these amendment requests.Concerning the above, Catawba surveillances with identical corresponding TSTF-425 surveillance numbers (item "a" above) are not deviations from TSTF-425.Catawba surveillance numbers that differ from the corresponding TSTF-425 surveillance numbers (item "b" above) are administrative deviations only from TSTF-425 with no impact on the NRC Staff's model SER.TSTF-425 surveillances that are not contained in the Catawba TS (item "c" above) are not applicable to these amendment requests.
This also includes Catawba's corresponding surveillances that are event driven or performed in accordance with an existing program (safety evaluation scope exceptions).
Not including these TSTF-425 surveillances is an administrative deviation from TSTF-425 with no impact on the NRC Staff's model SER.ForCatawba plant specific surveillances that are not contained in TSTF-425 Surveillance mark-ups, but are applicable to this amendment request (item "d" above), Duke Energy has determined that the relocation of these surveillance frequencies is consistent with TSTF-425, Revision 3, and the NRC Staff's model SER. This includes the scope exceptions documented in Section 1.0,"Introduction," of the model SER since the Catawba plant specific surveillances involve fixed periodic frequencies and therefore do not meet any of the four exceptions.
A similar cross reference table comparing the TSTF and plant specific surveillances was also provided by Exelon's Peach Bottom Atomic Power Station. The License Amendment applications to relocate specific surveillances in accordance with TSTF-425 are dated August 31, 2009 (NRC Accession No.ML092470153) and October 30, 2009 (NRC Accession No. ML093060126).
Page 2 of 3 Catawba Nuclear Station Attachment 1 Description and Assessment Duke Energy currently has seven license amendment requests that are pending approval from the NRC that affect surveillances modified in this amendment request. A listing of those amendment letters is provided in the table below, along with the surveillances (SRs) that are affected.
Since the approval process of these amendments is in progress, Duke Energy will not know the final disposition of each request until later in 2010. Duke Energy will provide updated pages and mark-ups for affected SRs before final approval of this amendment.
Date of Affected SRs and SR Bases Amendment Letter 09/02/08 This LAR modifies the SR 3.6.6.4 to be "not applicable".
The SR Description for SR 3.5.4.2 is modified.
The Bases only for SRs 3.3.2.7, 3.3.2.9 and 3.6.6.3 are revised, also.10/02/08 This LAR modifies the SR Description for SRs 3.6.13.1, 3.6.13.4, and 3.6.13.5 and SR 3.6.13.6.
'05/28/09 This LAR modifies the SR Description of SRs 3.8.1.2, 3.8.1.7, 3.8.1.9, 3.8.1.11, 3.8.1.12, 3.8.1.15, 3.8.1.19, and 3.8.1.20.07/01/09 The SR Bases for SRs 3.3.1.7 and 3.3.1.8 are modified.
Also, the SR Description for SR 3.3.1.11 is modified.09/30/09 This LAR modifies the Frequency of SR 3.6.6.7 to be event-driven.12/14/09 SRs 3.8.4.3 and 3.8.4.6 are modified to add a reference to a new table.12/15/09 SR 3.4.16.1 description is modified by this LAR. Also, SR 3.4.16.3 is deleted.3.0 Regulatory Analysis 3.1 No Significant Hazards Consideration Duke Energy has reviewed the proposed no significant hazards consideration determination (NSHC) published in the Federal Register on July 6, 2009, 74 FR'31996-32006. Duke Energy has concluded that the proposed NSHC presented in the Federal Register notice is applicable to Catawba Units 1 and 2, and is provided in Attachment 6 to this amendment request. This satisfies the requirements of 10 CFR 50.91 (a).Page 3 of 3 Attachment 2 Documentation of PRA Technical Adequacy Catawba Nuclear Station 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)
Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 Documentation of PRA Technical Adequacy TABLE OF CONTENTS Section Page 2 .1 O v e rv ie w .......................................................................................................................
2 2.2 Historical Summary ....... .................................
3 2.3 PRA Technical Adequacy Consistent With RG 1.200, Section 4.2 ...........................
4 2.3.1 PRA Model Adequately Represents the as-built, as-operated Plant ................
4 2.3.2 Unincorporated Changes to the Plant ..............................
6 2.3.3 Departuresfrom ASME Requirements
...............................................
6 2.3.4 M ethodology to be Used for Initiative 5b ...............................................................
6 2.3.5 Identification of Key Assumptions
.................................
.............................
7 2.3.6 Resolution of Relevant Peer Review/Self-Assessment Findings and O bserv ations .................................................................................................
.....7 2.3.7 Applicable Capability Category for Initiative 5b ....................................................
8 2.4 External Events Considerations
.................
......................
8 2.4.1 Overall External Hazards Analysis Methodology
.......................
8 2.4.2 Cataw ba Seism ic PRA M odel ..........................................................................
9 2.4.3 C ataw ba Fire PRA M odel ...............................................................................
11 2.4.3.1 Catawba Future State Fire PRA Model Initiative
.............................
12 2.4.4 Catawba Shutdown Risk Impact Analysis ......................................................
12 2 .5 S u m m a ry ......................................................................................................................
1 3 2 .6 R e fe re n c e s ..................................................................................................................
1 3 Table 2-1 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD THROUGH ADDENDA RA-Sc-2007
........ 16 Page 1 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 2.1 Overview The technical adequacy of the probabilistic risk assessment (PRA) must be compatible with the safety implications of the proposed Technical Specification (TS) changes and the role the PRA plays in justifying the changes. The Nuclear Regulatory Commission (NRC) has developed Regulatory Guide (RG) 1.200 (Ref. 1), which references the American Society of Mechanical Engineers (ASME) PRA standard RA-Sb-2005, Addenda to ASME RA-S-2002 (Ref. 2) for internal events at power. External events and shutdown risk impacts may be considered quantitatively or qualitatively.
RG 1.200 also references the NEI peer review process NEI 00-02 (Ref. 3).The industry guidance document for the implementation of Initiative 5b is NEI 04-10,"Risk-Informed Method for Control of Surveillance Frequencies".
The NRC issued a Final Safety Evaluation for NEI 04-10 Revision 0, on September 28, 2006 (Ref. 4). The Staff found that NEI 04-10, Revision 0, was acceptable for referencing by licensees proposing to amend their TSs to establish a Surveillance Frequency Control Program (SFCP), provided that the following conditions are satisfied:
- 1. The licensee submits documentation with regard to PRA technical adequacy consistent with the requirements of RG 1.200, Section 4.2.2. When a licensee proposes to use PRA models for which NRC-endorsed standards do not exist, the licensee submits documentation, which identifies the quality characteristics of those models, consistent with RG 1.200, Sections 1.2 and 1.3. Otherwise, the licensee identifies and justifies the methods to be applied for assessing the risk contribution for sources of risk not addressed by PRA models.Subsequently NEI 04-10 Revision 1 was approved (Ref. 5) and is the current document of record.The implementation of the SFCP at the Catawba Nuclear Station will follow the guidance provided in NEI 04-10, Revision 1 in evaluating proposed surveillance frequency changes.The Catawba PRA is a full scope PRA including both internal and external events (i.e., flood, seismic, fire, high winds (tornado)).
Having previously completed a self-assessment against the supporting requirements of ASME PRA Standard through addenda RA-Sc-2007 (Ref. 6), Duke Energy is planning to perform a self-assessment against the supporting requirements of ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 (Ref. 7) for the current Catawba PRA model of record (including fire, flood, seismic, and high winds (tornado) models) in 2010. Also there is currently significant work being performed at Duke Energy in the area of fire PRAs. This will be discussed further in the Fire PRA Model section.The following information is submitted by Duke Energy to address the conditions of the NRC SER for Initiative 5b.Page 2 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 2.2 Historical Summary The original Catawba PRA was initiated in July 1984 by Duke Power Company (Duke Power) staff, Delian Corporation, and Safety and Reliability Optimization Services (SAROS), Incorporated.
Law Engineering Testing Company and Structural Mechanics Associates provided specific input to the seismic analysis.
Science Applications International Corporation (SAIC) provided support for the human reliability analysis.External reviews were conducted by personnel from SAIC, Delian Corporation, and SAROS Incorporated.
It was a full scope Level 3 PRA with internal and external events (i.e., seismic, flood, high winds (tornado), fire): A peer review of the draft PRA was conducted by Electric Power Research Institute (EPRI) (Ref. 8). The final study, which incorporated the comments of all the reviews, was completed in August 1987 and resulted in an internal Duke Power report (Ref. 9) as Revision 0 to the PRA.On November 23, 1988, the NRC issued Generic Letter (GL) 88-20 (Ref. 10), which requested that licensees conduct an Individual Plant Examination (IPE) in order to identify potential severe accident vulnerabilities at their plants., In response, Duke Power initiated a review and update of the original Catawba study in April 1991. The Catawba response to GL 88-20 was provided by letter dated September 10, 1992 (Ref. 11). In this letter Duke Power noted that the enclosed Revision 1 of the PRA consisted of a complete Level 3 PRA with a detailed analysis of both internal and external events. By letter dated June 7, 1994 (Ref. 12), the NRC provided a SER of the internal events portion of the above Catawba IPE submittal.
In response to Generic Letter 88-20, Supplement 4 (Ref. 13), Duke Power completed an Individual Plant Examination of External Events (IPEEE) for severe accidents.
This IPEEE was submitted to the NRC by letter dated June 21, 1994 (Ref. 14). The IPEEE report contained a detailed write-up of the Catawba seismic and fire PRA analysis methods, results and conclusions.
It also addressed other events such as high winds, floods, and transportation accidents.
The IPEEE study did not identify any plant changes that would significantly reduce the risk from external events.Duke Power subsequently responded to an NRC Request for Additional Information (RAI) on the IPEEE submittal November 17, 1995 (Ref. 15). Duke Power also submitted a Supplemental IPEEE Fire Analysis Report to the NRC July 30, 1996 (Ref.16) in response to a request for supplemental fire investigations as described below.1. Develop fire accident sequences for those areas that were previously screened from further review because they were considered to be subsumed by other initiators.
- 2. Perform sensitivity studies on firedetection and suppression parameters.
- 3. Re-review cable routing to confirm that potential plant trip initiators have been considered in all areas.'The supplemental fire investigations in the report produced a more complete quantification of the fire induced core damage sequences but the conclusions and results were not significantly different from those reported in the original IPEEE report.Page 3 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 By letter dated April 12, 1999 (Ref. 17), the NRC provided an evaluation of the IPEEE submittal.
The conclusion of the NRC letter [page 6] states: "The staff finds the licensee's IPEEE submittal is complete with regard to the information requested by Supplement 4 to GL 88-20 (and associated guidance in NUREG-1407), and the IPEEE results are reasonable given the Catawba design, operation, and history. Therefore, the staff concludes that the licensee's IPEEE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities, and therefore, that the Catawba IPEEE has met the intent of Supplement 4 to GL 88-20." While 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, there have not been significant numbers of plant changes made since the initial NRC review that would invalidate the methodologies used in the existing external events models of record.In 1996, Catawba initiated Revision 2 of the 1992 IPE and provided the results to the NRC in 1998 (Ref. 18). Revision 3 of the Catawba PRA was completed in December 2004 and Revision 3a was completed in August 2006. Revision 3 was a major comprehensive revision to the PRA models and associated documentation.
Revision 3a was a change to implement the turbine building 4160 VAC transformer flood wall modifications and other various model enhancements.
Revision 3a is the current model of record. Work is currently underway on Revision 4 of the Catawba PRA, which is a major revision to the PRA, and includes a planned revision to the fire PRA model (discussed in Section 2.4.3.1).2.3 PRA Technical Adequacy Consistent With RG 1.200, Section 4.2 This section addresses Condition 1 of the NRC SER for Initiative 5b.2.3.1 PRA Model Adequately Represents the As-Built, As-Operated Plant The basis to conclude that the PRA model to be used adequately represents the as-built, as-operated plant is as follows.The existing PRA Configuration Control Program at Catawba was assessed against Section 5 of the ASME PRA Standard to meet the requirements necessary to support risk-informed decisions.
The results of the self-assessment concluded that the PRA fully meets the requirements for configuration control of a PRA to be used with the ASME PRA Standard to support risk-informed decisions for nuclear power plants. A summary of the program and the basis to conclude that the PRA model adequately represents the as-build, as-operated plant is provided below.The PRA Configuration Control Program at Catawba is governed by the following workplace procedures.
Page 4 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3* XSAA-101, Risk-Impact Review of Nuclear Plant Changes Including Nuclear Station Modification, and Emergency or Abnormal Procedure Changes* XSAA-106, PRA Maintenance, Update and Application XSAA-101 addresses the process for review of plant design changes, plant emergency and abnormal procedure changes, and Technical Specification (TS) changes that have been made for PRA impacts. It also describes in detail the process used to review the impact of potentially significant changes that could impact the PRA before the changes have been made.XSAA-106 addresses the conditions when a PRA update may be required (e.g., cumulative risk impact of unincorporated PRA changes exceeds a threshold such that the as-built as-operated plant is not adequately represented by the PRA). It addresses a process to assess the risk of a change to the plant and a method to prioritize the implementation of a plant change based on the risk impact to the PRA. It describes a process to ensure that an annual assessment is made of the cumulative impact of PRA changes that have not yet been incorporated into the PRA and provides guidance as to when a PRA update is needed based on the results of the annual assessment.
Finally, it describes the electronic tracking tool that is used to track changes that impact the PRA till they are incorporated into the PRA.The process requires notification of any completed (and planned changes that could significantly impact the PRA model) plant modifications, TS changes, or Emergency Procedure changes are sent to the PRA Section for a review of any PRA impacts. This review is documented.
If a plant change is determined to impact the PRA then is it entered into the electronic tracking tool where a risk assessment is performed on the change. The outcome of the risk assessment will "bucket" the plant change into a Low, Medium, or High risk change category based on the estimated delta Core Damage Frequency (CDF) or delta Large Early Release Frequency (LERF) results., Plant changes thatare determined to be of a Low risk impact are tracked to completion in the electronic tracking tool and are annually assessed for their cumulative impact on the PRA model. Plant changes that are determined to be of Medium or High risk impact are entered into the site corrective action program for further analysis as to their impact on current applications.
They also are tracked to completion in the electronic tracking tool and are annually assessed for their cumulative impact on the baseline PRA model.For any application that requires a PRA analysis (e.g., License Amendment Request (LAR) or Notice of Enforcement Discretion (NOED)) workplace procedures require that all of the outstanding PRA model changes listed in the electronic tracking tool are individually reviewed for their impact on the application.
A justification is made as to why each item does not impact the PRA results used to support the application.
This review is documented.
If it is determined that an unincorporated change might impact an application then steps are taken to either perform sensitivity studies to demonstrate that the contributors significant to the application were not impacted or the PRA model is revised to address the impact of the change on the application.
This analysis will also be performed and documented for every application of Initiative 5b.The outstanding items in the electronic tracking tool are ultimately incorporated into a major PRA revision which is performed periodically to ensure that the overall number of Page 5 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 items being tracked remains manageable.
This robust process, governed by written procedures, is sufficient to ensure the PRA model represents the as-built, as-operated plant.2.3.2 Unincorporated Changes to the Plant The justification of how unincorporated changes to the plant will be addressed is provided in the response in Section 2.3.1.2.3.3 Departures from ASME Requirements The justification for departures from the ASME Standard Capability Category II requirements, including any unresolved findings/observations is as follows.In March 2002, the PRA at Duke Energy's Catawba Nuclear Station received a peer review by an industry team of knowledgeable PRA practitioners (Ref. 19). Since the performance of this peer review, the industry has utilized the American Society of Mechanical Engineers (ASME) process to develop a standard identifying the requirements associated with PRA. RG 1.200 endorses the ASME PRA Standard as an acceptable method for demonstrating the technical adequacy of a PRA -provided various clarifications are made as identified in the regulatory guide.Subsequently in 2008, as noted earlier, Duke Energy conducted a self-assessment of the Catawba PRA (Ref. 6) against the ASME PRA Standard through addenda RA-Sc-2007.The Catawba PRA self-assessment included the Risk Assessment Technical Requirements listed in Section 4 of the ASME PRA Standard.
This self-assessment evaluated the PRA with respect to Capability Category I1. For the purposes of Initiative 5b, deviations from the Capability Category II supporting requirements were identified and dispositioned to ensure that these issues do not negatively impact Initiative 5b. For those requirements of the standard that have not been met, a justification of why it is acceptable that the requirement has not been met has been provided.
A summary of these items is shown in Table 2-1 for Catawba (Ref. 6). Of the 29 items, 26 are either documentation or have no expected impact on Initiative 5b applications.
The remaining three could have an impact based on the specific Initiative 5b application.
Because of the broad scope of potential Initiative 5b applications, and the fact that the impact of assumptions may differ for each surveillance requirement being evaluated, Duke Energy will address each of the deviations from Capability Category II listed in Table 2-1 for the Catawba PRA respectively for each application of Initiative 5b on an application specific basis. Again, if a requirement is not met a justification of why it is acceptable that the requirement has not been met will be provided.
These results will be with the documentation package for the specific Initiative 5b application.
2.3.4 Methodology to be Used for Initiative 5b Page 6 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 NEI 04-10 Revision 1 provides the detailed process requirements for controlling surveillance frequencies (SF) of the TS Surveillance Requirements (SRs) that have been relocated from the TSs to the SFCP. The methodology described in NEI 04-10 Revision 1 provides a risk-informed process to support a plant expert panel (called an Integrated Decisionmaking Panel or IDP) assessment of proposed changes to SF, assuring appropriate consideration of risk insights and other deterministic factors, which may impact SF, along with appropriate performance monitoring of changes and documentation requirements.
The Duke Energy SFCP, including the methodology of assessing SF.changes utilized at Catawba, is consistent with NEI 04-10, Revision 1 and the supporting background document TSTF-425-A, Rev. 3 (Ref. 20).2.3.5 Identification of Key Assumptions Identification of Key Assumptions related to SF (if any) and how they will be addressed is given below.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 a SF 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 SF change impact. Therefore, the methodology requires the performance of selected sensitivity studies on the standby failure rate of the component(s) of interest for the SF change assessment.
Because of the broad scope of potential Initiative 5b applications, any key assumptions and approximations relevant to the results obtained for an application of Initiative 5b will be addressed and documented on an application specific basis. This includes not only the results of the standby failure rate sensitivity study, but the results of any additional sensitivity studies identified during the performance of the reviews as outlined in Sections 2.3.1, 2.3.2, and 2.3.3.2.3.6 Resolution of Relevant Peer Review/Self-Assessment Findings and Observations Section 2.3.3 discusses departures from the ASME PRA Standard Capability Category II requirements and summarizes them on Table 2-1 for Catawba. However as previously noted, because of the broad scope of potential Initiative 5b applications, and the fact that the impact of assumptions may differ for each surveillance requirement being evaluated, Duke Energy will address each of the deviations from Capability Category II listed in Table 2-1 for Catawba for each application of Initiative 5b on an application'specific basis. If a requirement is not met a justification of why it is acceptable that the requirement has not been met will be provided.
If the PRA model is changed for a specific application of Initiative 5b to address self-assessment findings or if a sensitivity study is performed to demonstrate contributors significant to the application were not impacted by a self-assessment finding, a discussion of the results and conclusions for resolution will be included in the documentation package. Duke Energy will maintain a current listing of deviations from ASME PRA Standard Capability Category II requirements for Catawba for review and resolution against each application of Initiative 5b.Page 7 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 2.3.7 Applicable Capability Category for Initiative 5b In accordance with NEI 04-10 Revision 1, the PRA must meet Capability Category II to be used for Initiative 5b applications.
Duke Energy will ensure the Catawba PRA used for Initiative 5b applications either fully meets Capability Category II or departures from Capability Category II are justified to show insignificant impact on the results of the analysis.
This will be done by performing a review of all outstanding departures from Capability Category II against the specific Initiative 5b application being addressed.
The results of this review will be in the documentation package for the specific Initiative 5b application.
2.4 External Events Considerations This section addresses Condition 2 of the NRC Safety Evaluation for Initiative 5b.Specifically it identifies quality characteristics for PRA models for which NRC-endorsed Standards do not exist, consistent with RG 1.200, Sections 1.2 and 1.3, and justifies the methods to be applied for assessing the risk contribution for those sources of risk not addressed by PRA models.NRC endorsed standards currently exist for external hazards including seismic and fire PRAs. Revision 2 of Regulatory Guide (RG) 1.200 (Ref. 21), references the ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 (Ref. 7) for internal and external hazards. An NRC endorsed standard does not currently exist for shutdown PRAs. NEI 04-10 Revision 1 references RG 1.200 Revision 1 and ASME PRA Standard RA-Sb-2005b as the governing documents for Initiative 5b.The NEI 04-10 Revision 1 methodology allows for SF change evaluations to be performed in the absence of quantifiable PRA models for all external hazards. For those cases where the SF 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. In general, it is not expected that seismic, fire, or other external hazards will play a significant role in the impact of a given surveillance frequency change.This section discusses the Catawba overall external hazards analysis methodology, the Catawba specific seismic and fire PRAs, and describes the methodology to be used to address shutdown risk impacts for Initiative 5b consistent with the requirements of the NEI 04-10 Revision 1 methodology.
2.4.1 Overall External Hazards Analysis Methodology The general approach used to develop the external event PRA at Catawba is as follows: 1) Identify all natural and man-made credible external events that may affect the site using many reference sources.Page 8 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 2) A screening analysis was conducted using defined bounding criteria in order to select those events that may require further review.3) A scoping analysis was performed on the remaining non-screened events to determine those that warranted a detailed site and plant-specific analysis.This approach is consistent withthat previously submitted to the NRC in Section 2.3 of Reference 14 and Volume 1, Section 3.0 of Reference
- 11. These references provide a greater level of detail of the approach if needed.2.4.2 Catawba Seismic PRA Model As noted in the IPEEE submittal (Ref. 14), Catawba Unit 2 was selected for a trial assessment of the EPRI developed Seismic Margin Methodology, the methodology for assessing the ability of nuclear plants to withstand earthquakes beyond design basis.The assessment established that Catawba would survive earthquake loads up to approximately twice its design basis. This work is documented in EPRI NP-6359 (Ref.22).The current Catawba seismic PRA model of record was last updated as part of Revision 3 of the, PRA model (Ref. 23). However, the current methodology used is the same as that described in detail in the IPE submittal (Ref. 11) and Section 3 of the IPEEE submittal (Ref. 14) both of which have already been reviewed by the NRC. The reader is referred to those references for additional details of the seismic analysis.The plant-specific seismic PRA analysis consists of four steps each of which are described below: 1) The Catawba site was evaluated to obtain the seismic hazard in terms of the frequency of occurrence of ground motions of various magnitudes.
The site-specific hazard analysis (Ref. 24) was performed using the Seismicity Owners Group (SOG) methodology developed by EPRI for seismic hazard analysis of nuclear power plant sites in the Central and Eastern United States (CEUS).Uncertainties were addressed in the hazard analysis.2) From the site-specific seismic hazard curve, the capacities of important plant structures and equipment to withstand seismic events were evaluated to determine conditional probabilities of failure as a function of ground acceleration for significant contributors (i.e., SSCs). These are commonly referred to as 'fragilities' or the site-specific fragility curves. Plant walkdowns were conducted, the most recent ones consistent with the guidelines of EPRI NP-6041 (Ref. 25).3) An event tree was developed along with supporting top logic and system fault trees to reflect plant response to seismic events. These modified logic models were then solved to obtain Boolean expressions for the seismic event sequences of interest.4) The Boolean expressions were quantified by convolving the probabilistic site Page 9 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 seismicity and the fragilities for the plant structures and equipment obtained in steps 1 and 2. The resulting sequence frequencies are then integrated into the overall Catawba PRA risk results, resulting in final quantitative results.The major changes to the current seismic analysis that have been made since the IPEEE submittal are as follows: 1. Comprehensive review and revision of the seismic analysis documentation write-up.2. Added component/structure fragility information to support values used in analysis.3. Updated model with new Human Reliability Analysis (HRA) data.4. Updated model with new common cause data.5. Changes made to the fault tree are listed below.* Made a new top gate for the model to address containment safeguards responses.
Added all supporting logic for new containment safeguards responses gate. This change was made to aid in accident "binning" in the seismic analysis." After review reinstated components and structures back into the model that had previously been screened out. These include:; DC Charger> Refueling Water Storage Tank> 125V dc Battery Rack> Emergency Diesel Generator (EDG) Control Panel> Residual Heat Removal (RHR) Heat Exchanger> EDG Engine Control Panel> Neutral Ground Resistor> EDG Load Sequencer> Solid State Protection System (SSPS) Cabinets> 4160V ac Switchgear
> 125V dc Panelboard
- 120V ac Panelboard
> Inverter> Auctioneering Diode Assembly> EDG building> Control Rod Drive Mechanism Seismic Supports Auxiliary Building Shear Wall> 600V ac Motor Control Center* Added logic to include EDG seismic failures, loss of 120V ac, and EDG load sequencer failures.* Added Loss of Reactor Coolant (NC) Pump Seal Support logic to reflect the addition of a redesigned seal package. New seals are qualified for higher temperatures that limit the amount of leakage should failure occur." Miscellaneous logic additions to several systems to account for failures of component cooling as a support system.6. Updated the seismic event tree.7. Updated the seismic analysis quantitative results table.Page 10 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 As noted previously, Duke Energy is planning to perform a self-assessment against the supporting requirements for seismic events of ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 for the Catawba seismic PRA in 2010. The method as described in Section 2.3.3 of this attachment will be used to justify any departures from the ASME Standard Capability Category II requirements for each application of Initiative 5b. However, in accordance with the discussion in this section above, Duke considers the current seismic model of record as meeting the required quality characteristics of RG 1.200 Sections 1.2 and 1.3 and is therefore sufficient for use as is in the application of Initiative 5b SF changes.2.4.3 Catawba Fire PRA Model The current Catawba fire PRA model analysis and methodology (Ref. 26) used in the model of record is the same analysis and methodology as described in the IPE submittal (Ref. 11); Section 4 and Appendix B of the IPEEE submittal (Ref. 14); and as discussed in the Supplemental IPEEE Fire Analysis Report (Ref. 16), all of which have already been reviewed by the NRC. The reader is referred to those references for additional details of the fire analysis.The plant-specific fire PRA analysis consists of four steps each of which are described below: a. The Catawba site and plant areas were analyzed to determine critical fire areas and possible scenarios for the possibility of a fire causing one or more of a predetermined set of initiating events. Screening criteria were defined for those fire areas excluded from the fire analysis.b. If there was a potential for an initiating event to be caused by a fire in an area, then the area was analyzed for the possibility of a fire causing other events which would impact the ability to shutdown the plant. These were identified by reviewing the impact on the internal event analysis models.c. Each area was examined with an event tree fire model to quantify fire damage probabilities.
The event tree related fire initiation, detection suppression, and propagation probabilities to equipment damage states.d. Fire sequences were derived and quantified based on the fire damage probabilities and the additional failures necessary for a sequence to lead to a core melt. The additional failures were quantified by the models used in the internal events analysis.The major changes to the current fire analysis that have been made since the IPEEE submittal deal with implementation of changes from the Supplemental IPEEE Fire Analysis Report (Ref. 16) and revised base case fire initiating event frequencies.
Since the Catawba fire PRA model is integrated into the overall PRA model, quantitative fire risk insights will be obtained each time when the PRA model is exercised.
When the integrated PRA model is not utilized for a quantitative assessment and modeling of the Page 11 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 affected equipment is not feasible, the fire risk insights will be assessed qualitatively.
This approach is consistent with the accepted NEI 04-10 Revision 1 methodology.
Duke Energy is planning to perform a self-assessment against the supporting requirements for fire events of ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 for the Catawba fire PRA in 2010. The method as described in Section 2.3.3 of this attachment will be used to justify any departures from the ASME Standard Capability Category II requirements for each application of Initiative 5b. However, in accordance with the discussion in this section above, Duke Energy considers the current fire model of record as meeting the required quality characteristics of RG 1.200 Sections 1.2 and 1.3 and is therefore sufficient for use as is in the application of Initiative 5b SF changes.2.4.3.1 Catawba Future State Fire PRA Model Initiative In February 2005, Duke Energy notified the NRC (Ref. 27) of its intent to adopt National Fire Protection Association (NFPA) Standard 805, "NFPA 805, Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generation Plants," 2001 edition, pursuant to Section 50.48(c) of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50.48(c)), at all of its nuclear stations.In a letter dated June 8, 2005, the NRC accepted Duke Energy's intent to adopt 10 CFR 50.48(c) (NFPA 805 Rule) for all three sites with Oconee Nuclear Station beginning the transition as a pilot plant on June 1, 2005 (Ref. 28). Duke Energy was requested to inform the NRC when the transition would begin at Catawba.Subsequently, Duke Energy informed the NRC in 2007 (Ref. 29) that the transition to NFPA 805 at Catawba Nuclear Station had begun. The NRC response on January 4, 2008 (Ref. 30) acknowledged the transition to the performance-based standard for fire protection had begun at Catawba Units 1 and 2.The Catawba Fire PRA model being developed uses guidance contained in NUREG/CR-6850/EPRI TR-1011989, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities (Ref. 31). This is the same methodology and approach as that being used for the Oconee pilot. The Catawba Fire PRA model is to receive an industry peer review-against the requirements of Part 4 of ASME/ANS RA-Sa-2009, Addendum to RA-S-2008 (Ref. 7) in April 2010. When the'peer review report is received the departures from Capability Category II requirements and other findings will be addressed.
In September 2010, Duke Energy is planning to submit a License Amendment Request (LAR) to the NRC to adopt the new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a), 10 CFR 50.48(c), and the guidance in Regulatory Guide (RG) 1.205. A discussion of the peer review open items and their disposition is expected to be part of that submittal.
2.4.4 Catawba Shutdown Risk Impact Analysis Since no approved quantitative shutdown risk PRA model for shutdown events currently exists at Duke Energy, Catawba will either 1) utilize the plant shutdown safety assessment tool developed to support implementation of NUMARC 91-06 (Ref. 32) as Page 12 of 27 Catawba Nuclear Station Attachment 2-Adoption of TSTF-425, Revision 3 described in Duke Energy Nuclear Station Directive (NSD) 403 (Ref. 33) or 2) perform an alternate qualitative risk evaluation process to assess the proposed surveillance frequency change that utilizes Initiative 5b. These are acceptable options to not having a quantitative shutdown PRA model in accordance with Section 4 Step 10 (and other places) of NEI 04-10 Revision 1. In either case, the guidance of NEI 04-10 Revision 1 will be followed.2.5 Summary In Section 2.3 of this document the Catawba PRA technical adequacy was evaluated in accordance with the requirements of RG 1.200, Section 4.2. Section 2.4 of this document submitted quality characteristics of the seismic and fire PRA models in accordance with the requirements of RG 1.200, Sections 1.2 and 1.3. A discussion of the qualitative method to address shutdown risk was also discussed in Section 2.4.Because of the broad scope of potential Initiative 5b applications and the fact that the risk assessment details will differ from application to application, for each individual SF interval request, a review of the unincorporated changes to the plant and remaining gaps to specific requirements in the PRA standard will be made to determine which, if any, would merit additional application-specific sensitivity studies in the final analysis.The results of the discussions above provide a basis for concluding that the current Catawba Units 1 and 2 PRA model is sufficiently robust and suitable for use in risk-informed processes such as that proposed for the implementation of a Surveillance Frequency Control Program.2.6 References
- 1. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities", Revision 1, US Nuclear Regulatory Commission, January 2007.2. ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications", with Addenda ASME RA-Sa-2003 and ASME RA-Sb-2005, December 2005.3. NEI 00-02, "Probabilistic Risk Assessment Peer Review Process Guidance," Revision A3, Nuclear Energy Institute, March 20, 2000.4. Letter, USNRC to Nuclear Energy Institute, "Final Safety Evaluation for Nuclear Energy Institute (NEI) Industry Guidance Document NEI 04-10, Revision 0, "Risk-Informed Technical Specifications Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies"", September 28, 2006.5. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," April 2007.6. DPC-1535.00-00-0013 (Cross references:
CNC-1 535.00-00-0094, MCC-1535.00-00-0089, OSC-9380), "PRA Quality Self-Assessment, Catawba Units 1 & 2, McGuire Units I & 2, Oconee Units 1, 2 & 3", Revision 2, November 2009.7. ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.Page 13 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 8. "Catawba Nuclear Station, Unit 1 Probabilistic Risk Assessment, "Volume 1, Preface, Duke Power Company, August 1987.9. "Catawba Nuclear Station Unit 1 Probabilistic RiskAssessment," Volumes 1-3, Duke Power Company, August 18, 1987.10. NRC Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities", US Nuclear Regulatory Commission, November 23, 1988.11. Letter Duke Power Company to Document Control Desk (USNRC), Catawba Units 1 and 2, "Individual Plant Examination (IPE) Submittal in Response to Generic Letter 88-20," September 10, 1992.12. Letter USNRC to Duke Power Company, "Safety Evaluation of Catawba Nuclear Station, Units 1 and 2 Individual Plant Examination (IPE) Submittal," June 7, 1994.13. 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..14. Letter Duke Power Company to Document Control Desk (USNRC), Catawba Units 1 and 2, "Individual Plant Examination of External Events (IPEEE) Submittal," June 21, 1994.15. Letter Duke Power Company to Document Control Desk (USNRC), Catawba Nuclear Station, "Request for Additional Information-Individual Plant Examinations for External Events; Response," November 17, 1995.16. Letter Duke Power Company to Document Control Desk (USNRC), "Supplemental IPEEE Report," Duke Power Company, McGuire Nuclear Station, Catawba Nuclear Station, July 30, 1996.17: Letter USNRC to Duke Power Company, "Catawba Nuclear Station -Review of Individual Plant Examination of External Events (IPEEE)," April 12, 1999.18. Letter Duke Energy Corporation to Document Control Desk (USNRC), Catawba Units 1 and 2, "Probabilistic Risk Assessment (PRA), Revision 2 Summary Report, January 1998.19. "Catawba Nuclear Station Probabilistic Safety Assessment Peer Review Report", Westinghouse Electric Co. for the Westinghouse Owners Group, December 2002.20. Technical Specification Task Force Traveler number TSTF-425, Revision 3,"Relocate Surveillance Frequencies to Licensee Control -RITSTF Initiative 5," July 2009.21. Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities", Revision 2, US Nuclear Regulatory Commission, March 2009.22. EPRI NP-6359, "Seismic Margin Assessment of the Catawba Nuclear Station," April 1989.23. CNC-1535.00-00-0059, Catawba Nuclear Station, External Events -Seismic Analysis, December 2003.24. EPRI NP-4726-A, "Seismic Hazard Methodology for the Central and Eastern United States," July 1986.25. EPRI NP-6041, Revision 1, "A Methodology of Assessment of Nuclear Power Plant Seismic Margin, "August 1991.26. CNC-1535.00-00.0057, Catawba Nuclear Station, Fire Analysis Notebook, September 1997.27. Letter Duke Energy Corporation to Document Control Desk (USNRC), "Letter of Intent to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Page 14 of 27 Catawba Nuclear Station Attachment 2 Adoption of TSTF-425, Revision 3 Light Water Reactor Generating Plants, 2001 Edition," February 28, 2005 (Adams Accession Number ML050670305).
- 28. Letter USNRC to Duke Energy Corporation, "NRC Response to Duke's Letter of Intent to Adopt 10 CFR 50.48(c) (NFPA 805 Rule)," June 8, 2005 (Adams Accession Number ML051080005).
- 29. Letter Duke Energy Corporation to Document Control Desk (USNRC), "Letter of Intent to Start the Transition to NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants, 2001 Edition," June 7, 2007 (Adams Accession Number ML072260422).
- 30. Letter USNRC to Duke Energy Corporation, "Response to Letter of Intent toAdopt National Fire Protection Association Standard 805 for Duke Power Company's Catawba Nuclear Station, Units 1 and 2," January 4, 2008 (Adams Accession Number ML072780045).
- 31. EPRI/NRC-RES, "Fire PRA Methodology for Nuclear Power Facilities," NUREG/CR-6850, EPRI TR-1011989, Final Report, September 2005.32. NUMARC 91-06, "Guidelines for Industry Actions to Address Shutdown Management," December 1991.33. NSD-403, "Shutdown Risk Management (Modes 4, 5, 6, and No-Mode) per IOCFR 50.65(a)(4)," Revision 19, April 2009.Page 15 of 27 Catawba Nuclear Station Adoption of TSTF-425, Revision 3 Attachment 2 TABLE 2-1 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD THROUGH ADDENDA RA-Sc-2007 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application Gap #1 Accident sequence notebooks and AS-B3 Open. Phenomenological None -documentation system model notebooks should effects are considered in the issue.document the environmental effects model, although these of the initiating event and the impact considerations are not always on mitigation systems. documented.
Gap #2 Revise the data calc. to discuss DA-Ala Open. SSC and unavailability None -documentation component boundaries definitions.
boundaries, SSC failure modes issue.and success criteria are used consistently across analyses;however, these need to be formally documented.
Gap #3 Revise the data calc. to segregate DA-B1 Open. Previously, generic data Partitioning the failure standby and operating component sources often did not provide rates represents a data. Segregate components by standby and operating failure refinement to the data service condition to the extent rates. NUREG/CR-6928 does analysis process, but supported by the data. provided more of this data, and is not expected to will be used going forward. impact the 5b analysis.Gap #4 Enhance the documentation to DA-D4 Open. As part of the-Bayesian None -documentation include a discussion of the specific update process, checks are issue.checks performed on the Bayesian-performed to assure that the updated data, as required by this SR. posterior distribution is Page 16 of 27 Catawba Nuclear Station Adoption of TSTF-425, Revision 3 Attachment 2 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application reasonable given the prior distribution and plant experience.
These checks need to be formally documented.
Gap #5 Provide documentation of the DA-D6 Open. Generic common cause None -documentation comparison' of the component failure (CCF) probabilities are issue.boundaries assumed for the generic considered for applicability to CCF estimates to those assumed in the plant. CCF probabilities the PRA to ensure that these are consistent with plant boundaries are consistent.
experience and component boundaries, although the CCF documentation needs to be enhanced to discuss component boundaries.
Gap #6 Enhance the Human Reliability HR-A2 Open. Based on evaluations Relative to post-Analysis (HRA) to consider the using the EPRI HRA calculator, initiator human error potential for calibration errors, calibration errors that result in probabilities (HEPs), failure of a single channel are equipment random expected to fall in the low 10-3 failure rates and range. maintenance unavailability, calibration HEPs are not expected to contribute significantly to overall equipment unavailability.
Additionally, the next Page 17 of 27 Catawba Nuclear Station Adoption of TSTF-425, Revision 3 Attachment 2 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application revision of the PRA will incorporate the potential for calibration errors in the HRA.Thus there is no impact on the 5b analysis.Gap #7 Identify maintenance and calibration HR-A3 Open. Based on evaluations Relative to post-activities that could simultaneously using the EPRI HRA calculator, initiator HEPs, latent affect equipment in either different calibration errors that result in human error trains of a redundant system or failure of multiple channels are probabilities, diverse systems. expected to fall in the low equipment random failure rates and 10-S range. maintenance unavailability, calibration HEPs and misalignment of multiple trains of equipment are not expected to contribute significantly to overall equipment unavailability.
Thus there is no impact on the 5b analysis.Gap #8 Develop mean values for pre-initiator HR-D6 Open. Pre-initiator HEPs are The suggested data generally set to relatively high refinement is not Page 18 of 27 Catawba Nuclear Station Adoption of TSTF-425, Revision 3 Attachment 2 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application HEPs. screening values, which bound expected to have a the mean values. Even so, pre- significant impact on initiator HEPs are not the results. Thus significant contributors to risk. there is no impact on the 5b analysis.Gap #9 Document in more detail the HR-G3 Open. Performance shaping None -documentation influence of performance shaping- factors are accounted for in the issue.factors on execution human error development of human error probabilities.
probabilities, although detailed documentation is not always available for every HRA input.Gap #10 Enhance HRA documentation HR-G4 Open. Thermal Hydraulic (T/H) None documentation accordingly.
analyses, simulator runs and issue.operator interviews are used in developing the time available to complete operator actions. The time at which the cue to take action is received is specified in the HEP quantification.
However, the HRA documentation needs to be enhanced to provide a traceable path to all analysis inputs.Gap #11 Document a review of the human HR-G6 Open. HFEs are reviewed by None -documentation failure events (HFEs) and their final knowledgeable site personnel Page 19 of 27 Catawba Adoption Nuclear Station of TSTF-425, Revision 3 Attachment 2 Title Description of Gap Aplplicable Current Status / Comment Importance to 5b SRs Application HEPs relative to each other to to assure high quality. issue.confirm their reasonableness given However, this review needs to the scenario context, plant history, be better documented.
procedures, operational practices, and experience.
Gap #12 Develop mean values for post- HR-G9 Open. The use of mean values The 5b analysis will initiator HEPs. for HEPs instead of lower include a sensitivity probability median values can study to evaluate the affect the PRA results. use of different HEPs if the calculated risk is close to the threshold.
- Gap #13 Develop more detailed HR-H2 Open. Operator recovery None -documentation documentation of operator cues, actions are credited only if they issue.relevant performance shaping are feasible, as determined by factors, and availability of sufficient the procedural guidance, cues, manpower to perform the action. performance shaping factors and available manpower.
As noted for HR-G3, -G4, and -G6 above, the documentation of these considerations needs to be enhanced.Gap #14 Various enhancements to the IE-Al Open. No technical issues are None -documentation initiating events analysis identified, just a need to issues.documentation.
IE-A3a enhance the documentation.
Page 20 of 27 Catawba Nuclear Station Adoption of TSTF-425, Revision 3 Attachment 2 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application I E-A4 IE-A4a I E-A5 I E-A6 I E-A7 IE-B1 lE-B2 IE-B3 IE-D3 Gap #15 Various enhancements to the internal IF-B3 Open. Until the flooding flood analysis:
analysis is IF-C2c upgraded, the* Discuss flood mitigative features.
IF-C3 potential for flood-* Address the potential for spray, jet induced failures of impingement, and pipe whip IF-C3b SSCs will be failures.
assessed on a case-Provide more analysis of flood IF-E6b by-case basis.propagation flowpaths.
Address potential structural failure of doors or walls due to flooding loads and the potential for barrier unavailability.
Address potential Page 21 of 27 Catawba Nuclear Station Adoption of TSTF-425, Revision 3 Attachment 2 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application indirect effects.Enhance the documentation to address all of the SR details.Gap #16 Explicitly model Reactor Coolant LE-C6 Open. This issue affects No impact on the 5b System (RCS) depressurization for certain small LOCAs. analysis.small Loss of Coolant Accidents However, since the small LOCA (LOCAs) and perform the contribution to LERF is small, dependency analysis on the HEPs. there is no significant impact on the PRA results.Gap #17 Various enhancements to the LERF LE-G3 Open. None -documentation documentation.
issue.-LE-G4 LE-G5 LE-G6 Gap #18 Perform and document a comparison LE-F3 Open. Since Catawba and None -documentation of PRA results with similar plants. McGuire are sister plants, in issues.QU-D3 practice, their results are often compared.
Also, comparisons performed for MSPI and other programs help identify causes for significant differences.
However, to fully meet this SR, the model quantification documentation needs to be enhanced to provide a results Page 22 of 27 Catawba Nuclear Station Adoption of TSTF-425, Revision 3 Attachment 2 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application comparison.
Gap #19 Perform and document sensitivity LE-F2 Open. Perform and analyses to determine the impact of document sensitivity the assumptions and sources of QU-E4 analyses to model uncertainty on the results. determine the impact of the assumptions and sources of model uncertainty on the 5b analysis results.Gap #20 Expand the documentation of the QU-F2 Open. These SRs pertain to None -documentation PRA model results to address all QU-F6 the model quantification issues.required items. documentation.
Gap #21 Improve the documentation on the SC-A4 Open. Success criteria are None -documentation T/H bases for all safety function developed to address all of the issue.success criteria for all initiators, modeled initiating events.However, the documentation of success criteria needs to be improved to include initiator information.
Gap #22 Provide evidence that an SC-B5 Open. Catawba success None -documentation acceptability review of the T/H criteria are consistent with issue.analyses is performed.
those of sister plants included in the Pressurized Water Reactor Owners Group (PWROG) Probabilistic Safety Assessment (PSA) database.Page 23 of 27 Catawba Nuclear Station Adoption of TSTF-425, Revision 3 Attachment 2 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application However, to fully meet this SR, the success criteria documentation needs to be enhanced to include a results comparison.
Gap #23 Expand the documentation of the SC-Cl Open. These SRs pertain to None -documentation success criteria development to SC-C2 the success criteria issues.address all required items. documentation.
Gap #24 Enhance the system documentation SY-A4 Open. To support system None -documentation to include an up-to-date system model development, issue.walkdown checklist and system walkdowns and plant personnel engineer review for each system. interviews were performed.
However, documentation of an up-to-date system walkdown is not included with each system notebook.Gap #25 Enhance systems analysis SY-A8 Open. Basic event component None -documentation documentation to discuss component boundaries utilized in the issue.boundaries.
systems analysis are consistent with those in the data analysis.In addition, component boundaries are consistent with those defined in the generic failure rate source documents, such as NUREG/CR-6928.
Dependencies among components, such'as interlocks, are explicitly Page 24 of 27 Catawba Nuclear Station Adoption of TSTF-425, Revision 3 Attachment 2 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application modeled, consistent with the PRA Modeling Guidelines workplace procedure.
There is no evidence of a technical problem with component boundaries, just a need to improve the documentation.
Gap #26 Provide quantitative evaluations for SY-A14 Open. It is expected that There is no evidence screening.
conversion to a more of a technical problem quantitative approach would not associated with the change decisions about screening of whether or not to exclude components or components or failure modes. component failure A review of our qualitative modes, just a need to screening process confirms this document a expectation.
For example, quantitative screening.
transfer failure events for Thus there is no motor-operated valves (MOVs) impact on the 5b with 24 hr exposure times may analysis.not be modeled unless probabilistically significant with respect to logically equivalent basic events. For Catawba, the MOV transfers failure probability is less than 1% of the MOV fails to open on demand.failure rate. In cases like this, not including the Page 25 of 27 Catawba Adoption Nuclear Station of TSTF-425, Revision 3 Attachment 2 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application relatively low probability failure mode in the PRA model does not have an appreciable impact on the results.Gap #27 Per Duke Energy's PRA modeling SY-B8 Open. As noted for SY-A4, None -documentation guidelines, ensure that a walkdowns (which look for issue.walkdown/system engineer interview spatial and environmental checklist is included in each system hazards) have been performed, notebook.
Based on the results of the although up-to-date walkdown system walkdown, summarize in the documentation is not included system write-up any possible spatial with each system notebook.dependencies or environmental hazards that may impact system operation.
Gap #28 Document a consideration of SY-B15 Open. The impact of adverse None -documentation potential SSC failure due to adverse environmental conditions on issue.environmental conditions.
SSC reliability is considered but is not always documented.
However, there is no evidence of a technical problem associated with components that may be required to operate in conditions beyond their environmental qualification, just a need to improve the documentation.
Page 26 of 27 Catawba Nuclear Station Adoption of TSTF-425, Revision 3 Attachment 2 Title Description of Gap Applicable Current Status / Comment Importance to 5b SRs Application Gap #29 Enhance system model SY-C2 Open. This SR pertains to the None -documentation documentation to comply with all systems analysis issue.ASME PRA Standard requirements.
documentation.
1, j Page 27 of 27 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATION CHANGES INSERTS Insert I In accordance with the Surveillance Frequency Control Program REVIEWER'S NOTE: Text deleted and replaced by Insert I Will be relocated to the Surveillance Frequency Control Program (SFCP) document(s) per TSTF-425.Insert 3 5.5.17 Surveillance Freauency Control Program This Program provides controls for Surveillance Frequencies.
The program shall ensure that the Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operations are met.a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.Note: Insert 2 is included on Attachment
- 4.
Definitions 1.1 1.1 Definitions (continued)
RATED THERMAL POWER (RTP)REACTOR TRIP SYSTEM (RTS) RESPONSE TIME SHUTDOWN MARGIN (SDM)RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt.The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.
In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming: a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn.
With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA 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 zero power design level.SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing each slave relay and verifying the OPERABILITY of each slave relay. The SLAVE RELAY TEST shall include, as a minimum, a continuity check of associated testable actuation ,A STAGGERF_)
TEST BASIS shal of the testin g'f on"e of the s tems, subsystems/
hannels, or other /1 designate components durin e interval specifie y the Surveill 'ce Frequency, so t~at all systems, subs tems, /chan Is, or other designa/td components are. sted during/n S drveillance Frequend yintervals, where niche total .0Umber of systems, s usystems, channels zr other /designated componefts in the associate unction.THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.THERMAL POWER (continued)
Amendment Nos. Catawba Units 1 and 2 1.1-5 SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)LCO 3.1.1 APPLICABILITY:
SDM shall be within the limit specified in the COLR.MODE 2 with ke, < 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 limit specified in the COLR. 2 Catawba Units 1 and 2 3.1.1 -1 Amendment Nos. Gýo Core Reactivity 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1-------------
NOTE --------------------
The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.Verify measured core reactivity is within _ 1% Ak/k of predicted values.Once prior to entering MODE 1 after each refueling AND/2?1 Amendment Nos.(7 Catawba Units 1 and 2 3.1.2-2 Rod Group Alignment Limits 3.1.4 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6,hours associated Completion Time of Condition B not met.D. More than one rod not D.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 /> within alignment limit, limit specified in the COLR.OR D.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 /> required SDM to within limit.AND 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 /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual rod positions within alignment limit.AND Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter when the rod position deviation monitor is inoperable (continuea)
Catawba Units 1 and 2 3.1.4-3 Ame-,,ýHrnent NoSGý;iu Rod Group Alignment Limits 3.1.4 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.1.4.2 Verify rod freedom of movement (trippability) by moving each rod not fully inserted in the core > 10 steps in either direction.
SR 3.1.4.3 Verify rod drop time of each rod, from the fully withdrawn Prior to reactor position, is < 2.2 seconds from the beginning of decay of criticality after stationary gripper coil voltage to dashpot entry, with: each removal of the reactor head a. Tavg > 551°F; and b. All reactor coolant pumps operating.
Catawba Units 1 and 2 Amendment 3.1.4-4 Shutdown Bank Insertion Limits 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown bank is within the limits specified in the COLR.Catawba Units 1 and 2 Amendment Nos. 1 3.1.5-2 Control Bank Insertion Limits 3.1.6 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.1.6.2 Verify each control bank insertion is within the limits specified in the COLR.AND 5 C 17 Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter when the rod insertion limit monitor is inoperable SR 3.1.6.3 Verify sequence and overlap limits specified in the COLR are met for control banks not fully withdrawn from the core.Eli.Catawba Units 1 and 2 3.1.6-3 Amendment Nos.
PHYSICS TESTS Exceptions 3.1.8 ACTIONS (continued)
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 CHANNEL OPERATIONAL TEST on power Prior to initiation of range and intermediate range channels per SR 3.3.1.7, PHYSICS TESTS SR 3.3.1.8, and Table 3.3.1-1.SR 3.1.8.2 Verify the RCS lowest loop average temperature is [jndtE > 541 'F.SR 3.1.8.3 Verify THERMAL POWER is < 5% RTP. 0_SR 3.1.8.4 Verify SDM is within the limit specified in the COLR. K Catawba Units 1 and 2 3.1.8-2 Amendment FQ(X,Y,Z)3.2.1 SURVEILLANCE REQUIREMENTS
NOTE------------------------
During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify FQo(X,Y,Z) is within steady state limit.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 Fm(X,Y,Z) was last verified AND (continued)
Catawba Units 1 and 2 3.2.1-3 Amendment Nos.
Fo(X,Y,Z)3.2.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.2.1.2 --------------------
NOTE -------------------
- 1. Extrapolate FM(X,Y,Z) using at least two measurements to 31 EFPD beyond the most recent measurement.
If FM(X,Y,Z) is within limits and the 31 EFPD extrapolation indicates:
FQ(X,Y,Z)EXTRAPOLATED
> FQ(X,Y,Z)
EXTRAPOLATED, and EM)OKXY,Z)ExTRAPOLATED
> FL LOP ~,ZO FQ(X,Y,Z)
EXTRAPOLATED F0(X,Y,Z)then: a. Increase FM (X,Y,Z) by the appropriate factor specified in the COLR and reverify L (XyZOP;Fo(X,Y,Z)
< FQ(X,Y,Z) or b. Repeat SR 3.2.1.2 prior to the time at which FM(X,Y,Z)
< F.(X,Y,Z)°P is extrapolated to not be met.2. Extrapolation of Fm(X,Y,Z) is not required for the initial flux map taken after reaching equilibrium conditions.
Verify Fm(X,Y,Z)
-<_ FL(X,y,Z)°P.
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 FM(X,Y,Z) was last verified AND[5i1 ýi Catawba Units 1 and 2 3.2.1-4 Amendment Nos. 19ý t I =Uý12 FQ(X,Y,Z)3.2.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY+SR 3.2.1.3---- --NOTES -----------------
Extrapolate Fm(X,Y,Z) using at least two measurements to 31 EFPD beyond the most recent measurement.
If F"(X,Y,Z) is within limits and the 31 EFPD extrapolation indicates:
FM(X,Y,Z)EXTRAPOLATED F FL(X,Y,Z)RPSEXTRAPOLATED, and EMMXYZEXTRAPOLATED
> FM(X.Y.Z'F(X,Y,z)RPSEXTRAPOLATED F0(X,Y,Z)RPs then: a. Increase Fm(X,Y,Z) by the appropriate factor specified in the COLR and reverify Fm(X,Y,Z)
< FL(X,Y,Z)RPS; or b. Repeat SR 3.2.1.3 prior to the time at which F%(X,Y,Z)
_< F1(X,Y,Z)RPs is extrapolated to not be met.2. Extrapolation of Fm(X,Y,Z) is not required for the initial flux map taken after reaching equilibrium conditions.
Verify F"(X,Y,Z)
< F (X,Y,Z)RPS.
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 F"(X,Y,Z) was last verified AND aet reaferJ ,_L_ t ..m Catawba Units 1 and 2 3.2.1-5 AmnmetNos.~i FAH(X,Y)3.2.2 SURVEILLANCE REQUIREMENTS
NOTE-------------------------
During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify FmH (X,Y) is within steady state limit.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 F" (X,Y) was'last verified AND E D h e er jEFT (continued)
Amendment Nos(Catawba Units 1 and 2 3.2.2-3 FAH(X,Y)3.2.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY+SR 3.2.2.2-----------------
NOTES ------------------
- 1. Extrapolate FMH(X,Y) using at least two measurements to 31 EFPD beyond the most recent measurement.
If FMH(X,Y) is within limits and the 31 EFPD extrapolation indicates:
FMAHYET I OLTDýFL zvvSURV F~AH(,Y)EC1-RPOLAED>
AH(X,Y)ExORAPOLATEOD and L:MAH(X.YJEXTRAPOLATED
> e'AHMZJX FLH(X,Y)SURVEXTRAPOLATED FLAH(X,y)sURv then: a. Increase FmH(X,Y) by the appropriate factor specified in the COLR and reverify M<FAH(X,Y) -FLH(X,Y)SURV; or b. Repeat SR 3.2.2.2 prior to the time at which FMH(X,Y) < FLH(XY)suRv is extrapolated to not be met.2. Extrapolation of FMAH(X,Y) is not required for the initial flux map taken after reaching equilibrium conditions.
Verify FMH(X,Y) < FH(xY)suRv, 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 FAH(X,Y) was last verified AND Catawba Units 1 and 2 3.2.2-4 Amendment Nos.(g II 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 specified in the COLR.---------------------------
NOTE -------------------------
The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.APPLICABILITY:
MODE 1 with THERMAL POWER > 50% RTP.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits. A.1 Reduce THERMAL 30 minutes POWER to < 50% RTP.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD within limits for each OPERABLE excore channel.AND Once within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and every 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> thereafter with the AFD monitor alarm inoperable Catawba Units 1 and 2 3.2.3-1 Aa-,endii-jent No4;ý QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 --------------------
NOTES ----------------------
I .----------
- 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.
- 3. 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 exceeding 50% RTP.Verify QPTR is within limit by calculation.
AND Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter with the QPTR alarm inoperable
+SR 3.2.4.2-------------------------------
NOTES ---------------------------------
Only required to be performed if input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75% RTP.Verify QPTR is within limit using the movable incore detectors.
Catawba Units 1 and 2 3.2.4-4 Amendment Nos.L -
RTS Instrumentation 3-3-1 SURVEILLANCE REQUIREMENTS
I -----------------------
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. r SR 3.3.1.2----------
NOTES ---------------------------------
- 1. Adjust NIS channel if absolute difference is > 2%.2. 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 THERMAL POWER is > 15% RTP............................................................................
Compare results of calorimetric heat balance calculation to Nuclear Instrumentation System (NIS) chohnel output.+SR 3.3.1.3--------------------------
NOTES--------------------
- 1. Adjust NIS channel if absolute difference i~s > 3%.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 THERMAL POWER is > 15% RTP.('13 er EF)Compare results of the incore detector measurements to NIS AFD.(continued)
Catawba Units 1. and 2 3.3.1-9 Amendment Nos. ,
RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.1.4 This Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service.Perform TADOT.SR 3.3.1.5 Perform ACTUATION LOGIC TEST.SR 3.3.1.6-------- ------------------------
..I r~ -----------...
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 THERMAL POWER is > 75% RTP.Calibrate excore channels to agree with incore detector measurements.
11ýz 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 entry into MODE 3.Perform COT.(continued)
( A ' pS T B S " a s it p i o U i* The SR 3.3.1 .- equency of "days on a STAGG -ED TEST BASIS" as itapplies to Unit 2 Train 2A agnTrain 2B reapci trip breaker testin ay be extended on aroaicme basis to March 102009 at 0500 -rs, upon which Uni 2 shall be in Mode 3 wi~treactor trip brears open, r the End of C e 16 Refueling O ge. Upon entry into M6d4 3 with reactor tri br kers open for is refueling outage, his extension shall expird The provisions, R 3.0.2, rnot applica to this extension.,, -- -. -Catawba Units 1 and 2 3.3.1-10 Amendment Nos(ý)
RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.1.8 This Surveillance shall include verification that interlocks P-6 (for the Intermediate Range channels) and P-10 (for the Power Range channels) are in their required state for existing unit conditions.
Perform COT.i2 1-7---------
NOTE -------Only required when not*performed within Prior to reactor startup AND Four hours after reducing power below P-10 for power and intermediate range instrumentation AND Four hours after reducing power below P-6 for source range instrumentation AND (continued)
Catawba Units 1 and 2 3.3.1-11 Amendment Nos 424 RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.1.9----------------------------------
NOTE -------------------
Verification of setpoint is not required.Perform TADOT.&9 ýý: mcm SR 3.3.1.10 ----------------------
NOTE -------------------
This Surveillance shall include verification that the time constants are adjusted to the prescribed values.Pe....orm...CHANNEL....CALIBRATION........................................
Perform CHANNEL CALIBRATION.l[s.--
F SR 3.3.1.11--------------------
NOTE -------------------
- 1. Neutron detectors are excluded from CHANNEL CALIBRATION.
- 2. Power and Intermediate Range Neutron Flux detector plateau voltage verification is not required to be performed prior to entry into MODE 1 or 2.Perform CHANNEL CALIBRATION.
SR 3.3.1.12 Perform CHANNEL CALIBRATION.
SR 3.3.1.13 Perform COT.(continued)
Catawba Units 1 and 2 3.3.1-12 Amendment Nos. .
RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.1.14---------------------
NOTE -----------------
Verification of setpoint is not required.Perform TADOT.0 f6t ýh+SR 3.3.1.15-------------------
NOTE ..... ---------------------------
Verification of setpoint is not required.--------- NOTE -------Only required when not performed within previous 31 days Prior to reactor startup Perform TADOT.SR 3.3.1.16 -----...---
......-------------
NOTE --------------------------------
Neutron detectors are excluded from response time testing.Verify RTS RESPONSE TIME is within limits. 1 __-ns0 SR 3.3.1.17 Verify RTS RESPONSE TIME for RTDs is within limits. 11, _5s, Catawba Units 1 and 2 3.3.1-13 Amendment Nos. (1:7ýj n5, ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS
-NOTE ------------------------------------------------------------
Refer to Table 3.3.2-1 to determine which.SRs apply for each ESFAS Function.SURVEILLANCE FREQUENCY i SR 3.3.2.1 Perform CHANNEL CHECK.+SR 3.3.2.2 Perform ACTUATION LOGIC TEST.-NOTE+SR 3.3.2.3-- -- -- -- -- --- -- -- -- -- -N O T E --------------------------------
Final actuation of pumps or valves not required.Perform TADOT.SR 3.3.2.4 Perform MASTER RELAY TEST.E2~ hour_9-2 days on a STAGGERED TEST BAS 31 days.,!, days on a STAGGERED TEST BASI S da 18 onths for y estinghou AR and Po &Br ieldMDR laytypes SR 3.3.2.5 Perform COT.SR 3.3.2.6 Perform SLAVE RELAY TEST.I SR 3.3.2.7 Perform COT.31 day.jnl,-rrinu~d)
Catawba Units 1 and 2 3.3.2-10 Amendment No<
ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS
---------
NOTE --------------------------------
Refer to Table 3.3.2-1 to determine which SRs apply for each ESFAS Function.SR 3.3.2.6 Perform SLAVE RELAY TEST.Catawba Units 1 and 2 3.3.2-10 Amendment NosW ESFAS Instrumentation 3.3.2 SURVEILLANCE FREQUENCY SR 3.3.2.8-----------------
NOTE ................-
......
Verification of setpoint not required for manual initiation functions.
Perform TADOT.SR 3.3.2.9 ---------------------
NOTE- -----------------
This Surveillance shall include verification that the time constants are adjusted to the prescribed values.Perform....CHANNEL....CALIBRATION..........................................
P erform CHANNEL CALIBRATION.
/ #" SR 3.3.2.10------------------
NOTE --------------------------------
Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after SG 'pressure is > 600 psig.Verify ESFAS RESPONSE TIMES are within limit.1i8m ths ST ASI SR 3.3.2.11 Perform COT.=8,;wir SR 3.3.2.12 Perform ACTUATION LOGIC TEST.Catawba Units 1 and 2 Amendment Nos. [.3.3.2-1,1 PAM Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS N-------------------NC) r ---------------------------------------------------------------
SR 3.3.3.1 and SR 3.3.3.3 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 instrumentation channel that is normally energized.
SR 3.3.3.2 Not Used SR 3.3.3.3---------------
--NOTES ----------------
- 1. Neutron detectors are excluded from CHANNEL CALIBRATION.
- 2. CHANNEL CALIBRATION may consist of an electronic calibration of the Containment Area -High Range Radiation Monitor, not including the detector, for range decades above 10 R/h and a one point calibration check of the detector below 10 R/h with an installed or portable gamma source.'T Perform CHANNEL CALIBRATION.
Catawba Units 1 and 2 3.3.3-3 Amendment Nos. Gý?14 Remote Shutdown System 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform CHANNEL CHECK for each required instrumentation channel that is normally energized.
SR 3.3.4.2 .--- ------NOTE- -Not applicable to Reactor Trip Breaker Position.Perform CHANNEL CALIBRATION for each required 1 n instrumentation channel.Catawba Units 1 and 2 3.3.4-2 Amendment Nos4; ,
LOP DG Start Instrumentation 3.3.5 SURVEILLANCE REQUIREMENTS FREQUENCY SURVEILLANCE SR 3.3.5.1-NOTE --------------------------------
Testing shall consist of voltage sensor relay testing excluding actuation of load shedding diesel start, and time delay times.Perform TADOT.SR 3,3.5.2 Perform CHANNEL CALIBRATION with NOMINAL TRIP SETPOINT and Allowable Value as follows: a. Loss of voltage Allowable Value > 3242 V.Loss of voltage NOMINAL TRIP SETPOINT =3500 V.b. Degraded voltage Allowable Value > 3738 V.Degraded voltage NOMINAL TRIP SETPOINT 3766 V.Catawba Units 1 and 2 3-3.5-2 Amendment Nos.7 (n i~tt LOP DG Start Instrumentation B 3.3.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.5.2 r SR 3.3.5.2 is the performance of a CHANNEL CALIBRATION.
The setpoints, as well as the 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, as shown in Reference 1.ýap imately at every r eling. CHANNEL CALIBRATION is a compnete chec o e instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.The Frequency 18 months is b d on operat* g experience and consistency , h the typical ind try refueling cle and is justifi by the assumpti of an 18 month alibration inte a l in the determi tion of the Smagni de of equipmentift in the setp t analysi REFERENCES
- 1. UFSAR, Section 8.3.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.3.5-6 Revision No. ba ...
Containment Air Release and Addition Isolation Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS
NOTE---------------------------------
Refer to Table 3.3.6-1 to determine which SRs apply for each Containment Air Release and Addition Isolation Function..............................................................................................................................
SURVEILLANCE SR 3.3.6.1 Perform ACTUATION LOGIC TEST..,SR 3.3.6.2 Perform MASTER RELAY TEST.SR 3.3.6.3 Perform SLAVE RELAY TEST.SR 3.3.6.4 ---- -----------------
NOTE -----------------
Verification of setpoint is not required.Perform TADOT.Catawba Units 1 and 2 3.3.6-2 Amendment Nos 620)
BDMS 3.3.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4-SR 3.3.9.1 Perform CHANNEL CHECK.GýýSR 3.3.9.2 Perform COT. (3 ,a SR 3.3.9.3 Verify each automatic valve moves to the correct position and Reactor Makeup Water pumps stop upon receipt of an actual or simulated actuation signal.SR 3.3.9.4 -----------------------
NOTE -------------------
Only required to be performed when used to satisfy Required Action A.3 or B.3.Perform CHANNEL CHECK on the Source Range Neutron Flux Monitors.SR 3.3.9.5----------------------
NOTE-- -----------------
Only required to be performed when used to satisfy Required Action A.3 or B.3............................................................................
Verify combined flowrates from both Reactor Makeup Water Pumps are < the value in the COLR.Z3d ?11-SR 3.3.9.6--------------------
NOTE --------------------------------
Only required to be performed when used to satisfy Required Action A.3 or B.3.Perform COT on the Source Range Neutron Flux Monitors.Catawba Units 1 and 2 3.3.9-3 Amendment Nos.6ýE40)
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY-t SR 3.4.1.1 Verify pressurizer pressure is within limits.1L2 SR 3.4.1.2 Verify RCS average temperature is within limits.SR 3.4.1.3. Verify RCS total flow rate is within limits.SR 3.4.1.4 Perform CHANNEL CALIBRATION for each RCS total flow indicator.
Catawba Units 1 and 2 3-4-1 *-3 Amendment Nos. 15 RCS P/T Limits 3.4.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE --------C.1 Initiate action to restore Immediately Required Action C.2 parameter(s) to within shall be completed limits.whenever this Condition is entered. AND C.2 Determine RCS is Prior to entering Requirements of LCO acceptable for continued MODE 4 not met any time in other operation.
than MODE 1, 2, 3, or 4.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1-----------------
NOTE ------------------
Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.Verify RCS pressure, RCS temperature, and RCS heatup and cooldown rates are within limits.3 i tes A J--Catawba Units 1 and 2 3.4.3-2 Amendment Nos.7i65 RCS Loops -MODES 1 and 2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.4 RCS Loops -MODES 1 and 2 LCO 3.4.4 APPLICABILITY:
Four RCS loops shall be OPERABLE and in operation.
MODES 1 and 2.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of LCO 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 /> not met.SURVEILLANCE REQUIREMENTS SURVEILLANCE J FREQUENCY SR 3.4.4.1 Verify each RCS loop is in operation.
Amendment Nos.Catawba Units 1 and 2 3.4.4-1 RCS Loops -MODES 3 3.4.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loops are in operation.
I SR 3.4.5.2 Verify steam generator secondary side water levels are> 12% narrow range for required RCS loops.ff i Wýs 14 ý ýýI SR 3.4.5.3 Verify correct breaker alignment and indicated power are available to the required pumps that are not in operation.
Catawba Units 1 and 2 3.4.5-3 Amendment Nos(
RCS Loops -MODES 4 3.4.6 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. One RHR loop 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 /> OPERABLE.AND ALL RCS loops inoperable.
C. Both required RCS or C.1 Suspend operations that Immediately RHR loops inoperable, would cause introduction of coolant into the RCS with OR boron concentration less than required to meet SDM No RCS or RHR loop in of LCO 3.1.1 and maintain operation.
keff < 0.99.AND C.2 Initiate action to restore Immediately one loop to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RHR or RCS loop is in operation.
SR 3.4.6.2 Verify SG secondary side water levels are > 12% narrow range for required RCS loops.2s$~&1 urs QjVs§rf SR 3.4.6.3 Verify correct breaker alignment and indicated power are available to the required pump that is not in operation.
Catawba Units 1 and 2 3.4.6-2 Amendment Nos(ýýD
[RCS I 1,(,,0-. -MAO)-S 5. Loops Filled 3.4-7 ACTIONlS CONDITION REQUIRED ACTION COMPLETION TIME-t *7 A. One RHR loop inoperable.
AND Required SGs secondary side water levels not within limits.A.1 Initiate action to restore a second RHR loop to OPERABLE status.OR A.2 Initiate action to restore required SG secondary side water levels to within limits.Immediately Immediately 4-B. Required RHR loops inoperable.
B.1 Suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.AND Immediately Immediately B.2 Initiate action to restore one RHR loop to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3-4.7-1 Verify one RHR loop is in operation.
rs SR 3-4.7.2 Verify SG secondary side water level is > 12% narrow range in required SGs.SR 3.4.7.3 Verily correct breaker alignment and indicated power are S available to the required RHR pump that is not in operation.
Catawba Units 1 and 2 3.4.7-2 Amend ment No ýe ;
RCS Loops -MODES 5, Loops Not Filled 3.4.8 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Required 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 SDM No RHR loop in of LCO 3.1.1.operation.
AND B.2 Initiate action to restore Immediately one RHR loop to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one RHR loop is in operation.
SR 3.4.8.2 Verify correct breaker alignment and indicated power are available to the required RHR pump that is not in operation.
Catawba Units 1 and 2 3.4.8-2 Amendment Nos Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water level is < 92% (1656 ft 3).E~2~SR 3.4.9.2 Verify capacity of each required group of pressurizer heaters is > 150 kW.SR 3.4.9.3 Verify required pressurizer heaters are capable of being powered from an emergency power supply.Catawba Units 1 and 2 3.4.9-2 Amendment Nos. (7731;?')\ýfý Pressurizer PORVs.3.4.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F. (continued)
F.2 Restore one block valve to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> OPERABLE status if three block valves are inoperable.
AND F.3 Restore remaining block 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> valve(s) to OPERABLE status.G. Required Action and G.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 F not AND met.G.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 SR 3.4.11.1------------------------
NO)TE---Not required to be met with block valve closed in accordance with the Required Action of Condition B or E.Perform a complete cycle of each block valve.(continued)
Catawba Units 1 and 2 3.4.11-3 Amendment Nos.6 Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY i SR 3.4.11.2-------------------
NOTE ------------------
Required to be performed in MODE 3 or MODE 4 when the temperature of all RCS cold legs is > 200'F.Perform a complete cycle of each PORV.SR 3.4.11.3-------------------------------
NO TE -----------------
This SR is not applicable to valve NC-36B.IAi&~EdZf
{Verify the nitrogen supply for each PORV is OPERABLE by: a. Manually transferring motive power from the air supply to the nitrogen supply, b. Isolating and venting the air supply, and c. Operating the PORV through one complete cycle.Catawba Units 1 and 2 3.4.11-4 Amendment Nos.(
LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of two pumps (charging, safety injection, or charging and safety injection) are capable of injecting into the RCS.SR 3.4.12.2 Verify each accumulator is isolated.SR 3.4.12.3 Verify RHR suction isolation valves are open for each required RHR suction relief valve.SR 3.4.12.4 Verify PORV block valve is open for each required r PORV.SR 3.4.12.5 ------ ..-----------
NOTE ----------------------------
Not required to be met until 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 < 21 0°F.Perform a COT on each required PORV, excluding actuation.
SR 3.4.12.6 Perform CHANNEL CALIBRATION for each required PORV actuation channel.--
J.s2J L SR 3.4.12.7 Verify associated RHR suction isolation valves are open, with operator power removed and locked in removed position, for each required RHR suction relief valve.lip Catawba Units 1 and 2 3.4.12-5 Amendment Nos.r2 RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 ------------
--NOTES..---------.------
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 Only required to establishment of steady state operation.
be performed during steady 2. Not applicable to primary to secondary LEAKAGE. state operation Verify RCS Operational LEAKAGE within limits by Ds performance of RCS water inventory balance.SR 3.4.13.2 ----------------------
NOTE- ---------------
--- 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 Only required to establishment of steady state operation.
be performed--------------------------------------
during steady state operation Verify primary to secondary LEAKAGE is < 150 gallons rs per day through any one SG.Catawba Units 1 and 2 3.4.13-2 Amendment Nos. 2212 RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY*1-SR 3.4.14.1----------------
NOTES ------------------
- 1. Not required to be performed in MODES 3 and 4.2. Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.
- 3. 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.Verify leakage from each RCS PIV is equivalent to < 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure > 2215 psig and < 2255 psig.In accordance with the Inservice Test' m, and -AND Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months AND 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 or flow through the valve (continued)
Catawba Units 1 and 2 3.4.14-3 Amendment NosýF5 RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.14.2 Verify RHR system interlock prevents the valves from 18 being opened with a simulated or actual RCS pressure signal > 425 psig.Catawba Units 1 and 2 3.4.14-4 Amendment Nos C17,_
RCS Leakage Detection instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of the containment 1 rys ;atmosphere particulate radioactivity monitor.SR 3.4.15.2 Perform COT of the containment atmosphere particulate radioactivity monitor.SR 3.4.15.3 Perform CHANNEL CALIBRATION of the containment floor and equipment sump level monitors.SR 3.4.15.4 Perform CHANNEL CALIBRATION of the containment atmosphere particulate radioactivity monitor.th SR 3.4.15.5 Perform CHANNEL CALIBRATION of the containment ventilation unit condensate drain tank level monitor.+SR 3.4.15.6 Perform CHANNEL CALIBRATION of the incore instrument sump level alarm.18t Catawba Units 1 and 2 3.4.15-4 Amendment Nos.(
RCS Specific Activity 3.4.16 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Tag < 500°F.Time of Condition A not met.OR DOSE EQUIVALENT 1-131 in the unacceptable region of Figure 3.4.16-1.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY* SR 3.4.16.1 Verify reactor coolant gross specific activity < 100/E s.pcVgm.SR 3.4.16.2 ----.-----
NOTE --Only required to be performed in MODE 1.Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity < 1.0 pCi/gm.AND Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-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 (continued)
Catawba Units 1 and 2 3-4-16-2 7 Amendment Nos. C17 5)
RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.4.16.3-NOTE ---------------------
Not required to be performed until 31 days after a minimum of 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 />.Determine E from a sample taken in MODE 1 after a minimum of 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 />.Day s i,-V 5A J Catawba Units 1 and 2 3.4.16-3 Amendment Nos.
RCS Loops -Test Exceptions 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.17 RCS Loops-- Test Exceptions LCO 3.4.17 APPLICABILITY:
The requirements of LCO 3.4.4, "RCS Loops -MODES 1 and 2," may be suspended, with THERMAL POWER < P-7.MODES 1 and 2 during startup and PHYSICS TESTS.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. THERMAL POWER A.1 Open reactor trip breakers.
Immediately
> P-7.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify THERMAL POWER is < P-7.SR 3.4.17.2 Perform a COT for each power range neutron flux-low Prior to initiation of and intermediate range neutron flux channel, P-1 0, and startup and P-1 3. PHYSICS TESTS Catawba Units 1 and 2 3.4.17-1 Amendment Nos.e Accumulators 3.5.1-SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open.SR 3.5.1.2 Verify borated water volume in each accumulator is 2 r, -> 7630 gallons and < 8079 gallons.SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is -s> 585 psig and < 678 psig.SR 3.5.1.4 Verify boron concentration in each accumulator is within the limits specified in the COLR.AND_-NOTE-Only required to be performed for affected accumulators Once within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of> 75 gallons that is not the result of addition from the refueling water storage tank I_SR 3.5.1.5 Verify power is removed from each accumulator isolation valve operator when RCS pressure is > 1000 psig.Catawba Units 1 and 2 3.5.1-2 Amendment Nos.
ECCS -Operating 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify the following valves are in the listed position with power to the valve operator removed.a 9 tS- ý ý::_ :-:: ý_ I I Number NI 1 62A NI121A N1152B N1183B N1173A NI178B NIIOOB N1147B Position Open Closed Closed Closed Open Open Open Open Function SI Cold Leg Injection SI Hot Leg Injection SI Hot Leg Injection RHR Hot Leg Injection RHR Cold Leg Injection RHR Cold Leg Injection SI Pump Suction from RWST SI Pump Mini-Flow SR 3.5.2.2 Verify each ECCS manual, power operated, and d__automatic valve in the flow path, that is not locked, j__A sealed, or otherwise secured in position, is in the correct position.SR 3.5.2.3 Verify ECCS piping is full of water.SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance with flow point is greater than or equal to the required the Inservice developed head. Testing Program (continued)
Catawba Units 1 and 2 3.5.2-2 Amendment Nos.( 6 ECCS -Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each ECCS automatic 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 3.5.2.6 Verify each ECCS pump starts automatically on an actual or simulated actuation signal.SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, each position stop is in the correct position.Centrifugal Charging Safety Injection Pump Injection Throttle Pump Throttle Valve Number Valve Number N114 N1164 NI16 N1166 N118 NI168 N120 N1170 SR 3.5.2.8 Verify, by visual inspection, that the ECCS containment sump strainer assembly is not restricted by debris and shows no evidence of structural distress or abnormal corrosion.
Catawba Units 1 and 2 3.5.2-3 Amendment Nos. (8 23 RWST 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 Verify RWST borated water temperature is > 70°F and rs< 100°F.SR 3.5.4.2 Verify RWST borated water volume is > 363,513 gallons. -SR 3.5.4.3 Verify RWST boron concentration is within the limits 7a specified in the COLR.Catawba Units 1 and 2 3.5.4-2 Amendment Nos. 6 Seal Injection Flow 3.5.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.5.1 -------------------
NOTE -------------------
Not required to be performed until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the Reactor Coolant System pressure stabilizes at> 2215 psig and < 2255 psig.Verify manual seal injection throttle valves are adjusted 31 to give a flow within limit with centrifugal charging pump operating and the charging flow control valve full open.Catawba Units 1 and 2 3.5.5-2 Amendment Nos.
Containment Air Locks 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4-SR 3.6.2.1----------------
NOTES ----------------------------------
- 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 accordance with the Containment Leakage Rate Testing Program.In accordance with the Containment Leakage Rate Testing Program SR 3.6.2.2 Perform a pressure test on each inflatable air lock door seal and verify door seal leakage is < 15 sccm.SR 3.6.2.3 Verify only one door in the air lock can be o time.f, Catawba Units 1 and 2 3.6.2-5 Amendment Nos.4 Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each containment purge supply and exhaust isolation valves for the lower compartment and the upper compartment, instrument room, and the Hydrogen Purge System is sealed closed, except for one purge valve in a penetration flow path while in Condition E of this LCO.SR 3.6.3.2 Verify each Containment Air Release and Addition System isolation valve is closed, except when the valves are open for pressure control, ALARA or air quality considerations for personnel entry, or for Surveillances that require the valves to be open.SR 3.6.3.3 -------------------
NOTE-----------------------
Valves and blind flanges in high radiation areas may be verified by use of administrative controls.Verify each containment isolation manual valve and blind y flange that is located outside containment or annulus and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.(continued)
Catawba Units 1 and 2 3.6.3-5 Amendment Nos.
Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.3.4------------
NOTE-..-...
...........-
Valves and blind flanges in high radiation areas may be verified by use of administrative means.Verify each containment isolation manual valve and blind flange that is located inside containment or annulus and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days SR 3.6.3.5 Verify the isolation time of automatic power operated In accordance with containment isolation valve is within limits, the Inservice Testing Program-_SR 3.6.3.6 Perform leakage rate testing for Containment Purge In accordance with System, Hydrogen Purge System, and Containment Air the Containment Release and Addition System valves with resilient seals. Leakage Rate Testing Program SR 3.6.3.7 Verify each automatic containment isolation valve that is not locked, sealed or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.18 (continued)
Catawba Units 1 and 2 3.6.3-6 Amendment Nos. (E D I I 2-"C1A%,GES THIS KfACE-.-4 L Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.3.8 Verify the combined leakage rate for all reactor building In accordance with bypass leakage paths is < 0.07 La when pressurized to > the Containment 14.68 psig. Leakage Rate Testing Program Catawba Units 1 and 2 3.6.3-7 Amendment
.'os. 192/184 Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCO 3.6.4 APPLICABILITY:
Containment pressure shall be > -0.1 psig and < +0.3 psig.MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment pressure A.1 Restore containment 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> not 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.ý r <-i ýAn; tt: -Catawba Units 1 and 2-3.6.4-1 Amendment Nos.Oi Containment Air Temperature 3.6.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment upper compartment average air temperature is within limits.SR 3.6.5.2 Verify containment lower compartment average air temperature is within limits.Catawba Units 1 and 2 3.6.5-2 Amendment Nos. (2ýý Containment Spray System 3-6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray System LCO 3.6.6 APPLICABILITY:
Two containment spray trains shall be OPERABLE.MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One containment spray A-1 Restore containment spray 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> train inoperable, train to OPERABLE status.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. 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each containment spray manual, power operated, and automatic valve in the flow pathl that is not locked, sealed, or otherwise secured in position is in the correct position.(continued)
Catawba Units 1 and 2 3.6-6-1 Amendment Nos. G253 D248 Containment Spray System 3.6.6 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.6.2 Verify each containment spray pump's developed head at In accordance with the flow test point is greater than or equal to the required the Inservice developed head. Testing Program SR 3.6.6.3 Verify each automatic containment spray 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.ý1 ý8/ It ýh-ý0154 Sen.,+-SR 3.6.6.4 Verify each containment spray pump starts automatically on an actual or simulated actuation signal.SR 3.6.6.5 Verify that each spray pump is de-energized and prevented from starting upon receipt of a terminate signal and is allowed to start upon receipt of a start permissive from the Containment Pressure Control System (CPCS).SR 3.6.6.6 Verify that each spray pump discharge valve closes or is prevented from opening upon receipt of a terminate signal and is allowed to open upon receipt of a start permissive from the Containment Pressure Control System (CPCS).L/Y5~r 11 SR 3.6.6.7 Verify each spray nozzle is unobstructed.
Catawba Units 1 an d 2 3.6.6-2 Amendment Nos.a365 HSS 3.6.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.8.1 Operate each HSS train for > 15 minutes.SR 3.6.8.2 Verify the fan motor current is < 69 amps when the fan speed is > 3560 rpm and < 3600 rpm with the hydrogen skimmer fan operating and the motor operated suction valve closed.y Ti SR 3.6.8.3 Verify the motor operated suction valve opens automatically and the fans receive a start permissive signal.SR 3.6.8.4 Verify each HSS train starts on an actual or simulated actuation signal after a delay of > 8 minutes and < 10 minutes.EF9ý6- IýW--T Catawba Units 1 and 2 3.6.8-2 Amendment Nos.(1715 HIS 3.6.9 Catawba Units 1 and 2.3.6.9-2 Amendment No<424 AVS 3.6.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.10.1 Operate each AVS train for > 10 continuous hours with heaters operating.
~j~~RrI SR 3.6.10.2 Perform required AVS filter testing in accordance with the In accordance with Ventilation Filter Testing Program (VFTP). the VFTP SR 3.6.10.3 Verify each AVS train actuates on an actual or simulated actuation signal.SR 3.6.10.4 Verify each AVS filter cooling bypass valve can be opened.18 ths~7~f~j~ri SR 3.6.10.5 Verify each AVS train flow rate is > 8100 cfm and < 9900 cfm.SR 3.6.10.6 Verify each AVS train produces a pressure equal to or more negative than -0.88 inch water gauge when corrected to elevation 564 feet.Catawba Units 1 and 2 3.6.10-2 Amendment Nos(2$
ARS 3.6.11 3.6 CONTAINMENT SYSTEMS 3.6.11 Air Return System (ARS)LCO 3.6.11 Two ARS trains shall be OPERABLE.APPLICABILITY:
MODES 1, 2,3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One ARS train A.1 Restore ARS train to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable.
OPERABLE status.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 3.6.11.1 Verify each ARS fan starts on an actual or simulated actuation signal, after a delay of > 8.0 minutes and< 10.0 minutes, and operates for > 15 minutes.(continued)
Catawba Units 1 and 2 3.6.11-1 Amendment Nos 1 ARS 3.6.11 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.11.2 Verify, with the ARS air return fan damfper closed and with the bypass dampers open, each ARS fan current is < 59.0 amps when the fan speed is > 1174 rpm and < 1200 rpm.SR 3.6.11.3 Verify, with the ARS fan not operating, each ARS motor 9 y operated damper opens automatically on an actual or simulated actuation signal after a delay of > 9 seconds and < 11 seconds.SR 3.6.11.4 Verify the check damper is open with the ARS fan operating.
SR 3.6.11.5 Verify the check damper is closed with the ARS fan not operating.
SR 3.6.11.6~43E~r~Verify that each ARS fan is de-energized or is prevented from starting upon receipt of a terminate signal from the Containment Pressure Control System (CPCS) and is allowed to start upon receipt of a start permissive from the CPCS.+SR 3.6.11.7 Verify that each ARS fan motor-operated damper is prevented from opening in the absence of a start permissive from the Containment Pressure Control System (CPCS) and is allowed to open upon receipt of a start permissive from the CPCS.Catawba Units 1 and 2 3.6.11-2 Amendment No. 7Z Amendment No. C66(nit 2))
Ice Bed 3.6.12 3.6 CONTAINMENT SYSTEMS 3.6.12 Ice Bed LCO 3.6.12 The ice bed shall be OPERABLE.APPLICABILITY:
MODES 1, 2,3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Ice bed inoperable.
A.1 Restore ice bed to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> OPERABLE status.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 3.6.12.1 Verify maximum ice bed temperature is < 27 0 F.(continued)
Amendment Nos. 65 Catawba Units 1 and 2 3-6.12-1 Ice Bed 3.6.12 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.12.2 --------------
NOTE -.--------------
The chemical analysis may be performed on either the liquid solution or on the resulting ice.Verify, by chemical analysis, that ice added to the ice Each ice addition condenser meets the boron concentration and pH requirements of SR 3.6.12.7.SR 3.6.12.3 Verify, by visual inspection, accumulation of ice on n structural members comprising flow channels through the ice bed is < 15 percent blockage of the total flow area for each safety analysis section.SR 3.6.12.4 Verify total mass of stored ice is _> 2,132,000 lbs by 1 calculating the mass of stored ice, at a 95 percent confidence, in each of three Radial Zones as defined below, by selecting a random sample of > 30 ice baskets in each Radial Zone, and Verify: 1. Zone A (radial rows 8, 9), has a total mass of_> 324,000 lbs 2. Zone B (radial rows 4, 5, 6, 7), has a total mass of>1,033,100 lbs 3. Zone C (radial rows 1, 2, 3), has a total mass of> 774,900 lbs SR 3.6.12.5 Verify that the ice mass of each basket sampled in SR 3.6.12.4 is _ 600 lbs.(continued)
Catawba Units 1 and 2 3.6.12-2 Amendment Nos4 2 Ice Bed 3.6.12 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY 4 SR 3.6.12.6 Visually inspect, for detrimental structural wear, cracks, corrosion, or other damage, two ice baskets from each group of bays as defined below: a. Group 1 -bays 1 through 8;b. Group 2 -bays 9 through 16; and c. Group 3 -bays 17 through 24.40 ths 54 nth SR 3.6.12.7---------------
NOTE ----------------..---------------
The requirements of this SR are satisfied if the boron concentration and pH values obtained from averaging the individual sample results are within the limits specified below.Verify, by chemical analysis of the stored ice in at least one randomly selected ice basket from each ice condenser bay, that ice bed: a. Boron concentration is > 1800 ppm and < 2330 ppm; and b. pH is > 9.0 and < 9.5.Catawba Units 1 and 2 3.6.12-3 Amendment Nos( O3 Ice Condenser Doors 3.6.13 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Restore ice condenser door 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> associated Completion to OPERABLE status and Time of Condition B not closed positions.
met.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 or C AND 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 /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.13.1 Verify all inlet doors indicate closed by the Inlet Door Position Monitoring System.SR 3.6.13.2 Verify, by visual inspection, each intermediate deck door is closed and not impaired by ice, frost, or debris.SR 3.6.13.3 Verify, by visual inspection, each top deck door: a. Is in place; and b. Has no condensation, frost, or ice formed on the door that would restrict its opening.(continued)
Catawba Units 1 and 2 3.6.13-2 Amendment Nos. GýD Ice Condenser Doors 3.6.13 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.13.4 Verify, by visual inspection, each inlet door is not impaired by ice, frost, or debris.SR 3.6.13.5 Verify torque required to cause each inlet door to begin 1 to open is < 675 in-lb. 1-SR 3.6.13.6 Perform a torque test on inlet doors.1 t SR 3.6.13.7 Verify for each intermediate deck door: a. No visual evidence of structural deterioration;
- b. Free movement of the vent assemblies; and c. Free movement of the door.Catawba Units 1 and 2 3.6.13-3 Amendment Nos(17765ý)
Divider Barrier Integrity 3.6.14 ACTIONS (continued)
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 not met. AND 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 /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.14.1 Verify, by visual inspection, all personnel access doors Prior to entering and equipment hatches between upper and lower MODE 4 from containment compartments are closed. MODE 5 SR 3.6.14.2 Verify, by visual inspection, that-the seals and sealing surfaces of each personnel access door and equipment hatch have: a. No detrimental misalignments;
- b. No cracks or defects in the sealing surfaces; and c. No apparent deterioration of the seal material.Prior to final closure after each opening AND----------
NOTE -------Only required for seals made of resilient materials SR 3.6.14.3 Verify, by visual inspection, each personnel access door After each or equipment hatch that has been opened for personnel opening transit entry is closed.(continued)
Catawba Units 1 and 2 3.6.14-2 Amendment Nos GýD Divider Barrier Integrity 3.6.14 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.14.4 Remove two divider barrier seal test coupons and verify both test coupons' tensile strength is > 39.7 psi.6i~I~2~SR 3.6.14.5 Visually inspect > 95% of the divider barrier seal length, and verify: a. Seal and seal mounting bolts are properly installed; and b. Seal material shows no evidence of deterioration due to holes, ruptures, chemical attack, abrasion, radiation damage, or changes in physical appearance.
Catawba Units 1 and 2 3.6.14-3 Amendment Nos.6ýID Containment Recirculation Drains 3.6.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE REQUIREMENTS FREQUENCY SR 3.6.15.1 Verify, by visual inspection, that: Prior to entering MODE 4 from a. Each refueling canal drain valve is locked open; MODE 5 after and each partial or complete fill of the b. Each refueling canal drain is not obstructed by canal debris.SR 3.6.15.2 Verify, by visual inspection that no debris is present in the upper compartment or refueling canal that could obstruct the refueling canal drain.ý26 1 m th SR 3.6.15.3 Verify for each ice condenser floor drain that the: a. Valve opening is not impaired by ice, frost, or debris;b. Valve seat shows no evidence of damage;c. Valve opening force is < 66 Ib; and d. Drain line from the ice condenser floor to the lower compartment is unrestricted.
Catawba Units 1 and 2 3.6.15-2 Amendment Nos.(16 Reactor Building 3.6.16 3.6 CONTAINMENT SYSTEMS 3.6.16 Reactor Building LCO 3.6.16 The reactor building shall be OPERABLE.APPLICABILITY:
MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor building A.1 Restore reactor building to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> inoperable.
OPERABLE status.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 3.6.16.1 Verify the door in each access opening is closed, except 3 a "s when the access opening is being used for normal transit entry and exit.(continued)
Catawba Units 1 and 2 3.6.16-1 Amendment Nos. 6yD Reactor Building 3.6.16 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.6.16.2 Verify that during the annulus vacuum decay test, the vacuum decay time is > 87 seconds.SR 3.6.16.3 Verify reactor building structural integrity by performing a visual inspection of the exposed interior and exterior surfaces of the reactor building.Catawba Units 1 and 2 3.6.16-2 Amendment Nos 6"f" SG PORVs 3.7.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify one of the nitrogen bottles on each SG PORV is pressurized
> 2100 psig.SR 3.7.4.2 Verify one complete cycle of each SG PORV.IA / ýe4 18 rgn h SR 3.7.4.3 Verify one complete cycle of each SG PORV block valve.Catawba Units 1 and 2 3.7.4-2 Amendment Nos. 6ýýO AFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1--------------------
NOTE ----------------------------
Not applicable to automatic valves when THERMAL POWER is < 10% RTP.Verify each AFW manual, power operated, and automatic valve in each water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.L SR 3.7.5.2----------
NOTE---------------
Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after > 600 psig in the steam generator.
Verify the developed head of each AFW pump at the flow test point is greater than or equal to the required developed head.In accordance with the Inservice Testing Program SR 3.7.5.3--------------------
NOTE---------------
Not applicable in MODE 4 when steam generator is relied upon for heat removal.Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.Elý -e-I&-A- I (continued)
Amendment Nos.Catawba Units 1 and 2 3.7.5-3 AFW System 3.7.5 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.7.5.4-----------------
NOTES-----------------
- 1. Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after > 600 psig in the steam generator.
- 2. Not applicable in MODE 4 when steam generator is relied upon for heat removal.Verify each AFW pump starts automatically on an actual or simulated actuation signal.E ý ýl ki -ý:1-ýSR 3.7.5.5 Verify proper alignment of the required AFW flow paths Prior to entering by verifying flow from the condensate storage system to MODE 2, each steam generator.
whenever unit has been in MODE 5 or 6 for> 30 days Catawba Units 1 and 2 3.7.5-4 Amendment Nos. (17b3/65 CSS 3.7.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify the CSS inventory is > 225,000 gal. 2, Catawba Units 1 and 2 3.7.6-2 Amendment Nos. GýD CCW System 3.7.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 1~SR 3.7.7.1--------------------------------------
NOTE ----------------------------
Isolation of CCW flow to individual components does not render the CCW System inoperable.
Verify each CCW manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.SR 3.7.7.2 Verify each CCW automatic valve in the flow path servicing safety related equipment that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.SR 3.7.7.3 Verify each CCW pump starts automatically on an actual 18 n or simulated actuation signal.Catawba Units 1 and 2 3.7.7-2 Amendment NSWS 3.7.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.8.1---------------------
NOTE---------------
Isolation of NSWS flow to individual components does not render the NSWS inoperable.
Verify each NSWS manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.SR 3.7.8.2-------------------
NOTE ------------------
Not required to be met for valves that are maintained in position to support NSWS single supply header operation.
Verify each NSWS automatic 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.18 o h dý-EKýrslnlcý-
D-SR 3.7.8.3 Verify each NSWS pump starts automatically on an actual or simulated actuation signal.Catawba Units 1 and 2 3.7.8-3 Amendment Nos.41 2 SNSWP 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Standby Nuclear Service Water Pond (SNSWP)LCO 3.7.9 APPLICABILITY:
The SNSWP shall be OPERABLE.MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SNSWP inoperable.
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 /> 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 3.7.9.1 Verify water level of SNSWP is > 571 ft mean sea level.SR 3.7.9.2--NOTE Only required to be performed during the months of July, August, and September.
Verify average water temperature of SNSWP is < 95°F at an elevation of 568 ft. in SNSWP.~r,.+SR 3.7.9.3 Verify, by visual inspection, no abnormal degradation, erosion, or excessive seepage of the SNSWP dam.Catawba Units I and 2 3.7.9-1 Amendment Nos.i2 8 CRAVS 3.7.10 REQU IRED -ACTIONS (fntinf-ued)-
CONDITION REQUIRED ACTION COMPLETION TIME G. One or more CRAVS G.1 Restore CRAVS train(s) 7 days train(s) heater heater to OPERABLE inoperable, status.OR G.2 Initiate action in 7 days accordance with Specification 5-6.6.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CRAVS train for > 10 continuous hours with the heaters operating.
?M;Uj%6ý 4%_SR 3.7.10.2 Perform required CRAVS filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). VFTP SR 3.7.10.3 Verify each CRAVS train actuates on an actual or simulated actuation signal.uqgýýp SR 3.7.10.4 Perform required CRE unfiltered air inieakage testing in In accordance with accordance with the Control Room Envelope Habitability the Control Room Program. Envelope Habitability Program Catawba Units 1 and 2 3.7.10-3 Amendment Nos6ý CRACWS 3.7.11 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. Two CRACWS trains D.1 Suspend movement of Immediately inoperable in MODE 5 recently irradiated fuel or 6, or during assemblies.
movement of recently irradiated fuel assemblies.
E. Two CRACWS trains E.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2, 3, or4.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.11.1 Verify the control room temperature is < 90'F. ur Catawba Units 1 and 2 3.7.11-2 Amendment Nos. 3F ABFVES 3.7.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Operate each ABFVES train for > 10 continuous hours 3 d with the heaters operating.
SR 3.7.12.2 Perform required ABFVES filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.12.3 Verify each ABFVES train actuates on an actual or simulated actuation signal.SR 3.7.12.4 Verify one ABFVES train can maintain the ECCS pump rooms at negative pressure relative to adjacent areas.Catawba Units 1 and 2 3.7-12-2 Amendment Nos.(17 L65)
FHVES 3.7.13 SR 3.7.13.3 Perform required FHVES filter testing in accordance with the Ventilation Filter Testing Program (VFTP).In accordance with the VFTP SR 3.7.13.4 Verify one FHVES train can maintain a pressure< -025 inches water gauge with respect to atmospheric pressure during operation at a flow rate < 36,443 cfm.mon on a ITAG REP TEBA SR 3.7.13.5 Verify each FHVES filter bypass damper can be closed. 1 Catawba Units 1 and 2 3.7.13-2 Amendment No. 176 = 9n Amendment No. 68(Unit 2)
Spent Fuel Pool Water Level 3.7.14 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Water Level LCO 3.7.14 The spent fuel pool 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 spent fuel pool.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pool water A.1 -----NOTE-------
level not within limit. LCO 3.0.3 is not applicable.
Suspend movement of Immediately irradiated fuel assemblies in the spent fuel pool.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the spent fuel pool water level is > 23 ft above the 7 a top of the irradiated fuel assemblies seated in the storage W%_. .racks.Catawba Units 1 and 2 3.7.14-1 Amendment Nos.(57165 Spent Fuel Pool Boron Concentration 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Boron Concentration LCO 3.7.15 The spent fuel pool boron concentration shall be within the limit specified in the COLR.APPLICABILITY:
When fuel assemblies are stored in the spent fuel pool.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pool boron .------------
NOTE ----------
concentration not within LCO 3.0.3 is not applicable.
lim it. ---------------------------------------------
A.1 Suspend movement of fuel Immediately assemblies in the spent fuel pool.AND A.2 Initiate action to restore Immediately spent fuel pool boron concentration to within limit.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the spent fuel pool boron concentration is within limit.Catawba Units 1 and 2 3.7.15-1 Amendment Nos.6 Secondary Specific Activity 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Secondary Specific Activity LCO 3.7.17 The specific activity of the secondary coolant shall be < 0.10 pCi/gm DOSE EQUIVALENT 1-131.APPLICABILITY:
MODES 1, 2, 3, and 4.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Specific activity not 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 /> within 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 3.7.17.1 Verify the specific activity of the secondary coolant is 3 y< 0.10 pCi/gm DOSE EQUIVALENT 1-131.Catawba Units 1 and 2 3.7.17-1 Amendment Nos.(
AC Sources -Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power availability for each offsite circuit.SR 3.8.1.2 --------------------
NOTES-----------------
- 1. Performance of SR 3.8.1.7 satisfies this SR.2. All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.3. A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR as recommended by the manufacturer.
When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.Verify each DG starts from standby conditions and achieves steady state voltage > 3740 V and < 4580 V, _and frequency
> 58.8 Hz and < 61.2 Hz.(continued)
Catawba Units 1 and 2 3.8.1-5 Amendment Nos(E AC Sources -Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.1.3 ---------------------
NOTES ----------------
- 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 3.8.1.7.Verify each DG is synchronized and loaded and operates for > 60 minutes at a load > 5600 kW and < 5750 kW.SR 3.8.1.4 Verify each day tank contains > 470 gal of fuel oil.SR 3.8.1.5 Check for and remove accumulated water from each day 3 tank.SR 3.8.1.6 Verify the fuel oil transfer system operates to transfer fuel s oil from storage system to the day tank.(continued)
Catawba Units 1 and 2 3.8.1-6 Amendment 2Nos AC Sources -Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.1.7 -----------------------
NOTE ---------------
All DG starts may be preceded by an engine prelube period.Verify each DG starts fromstandby condition and achieves in < 11 seconds voltage of > 3740 V and frequency of > 57 Hz and maintains steady-state voltage> 3740 V and < 4580 V, and frequency
> 58.8 Hz. and< 61.2 Hz.SR 3.8.1.8 Verify automatic and manual transfer of AC power h -sources from the normal offsite circuit to each alternate offsite circuit.(continued)
Catawba Units 1 and 2 3.8.1-7 Aatwmendment Nos.-
AC Sources -Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.1.9--------------------
NOTE ---------------
If performed with the DG synchronized with offsite power, it shall be performed at a power factor < 0.9.Verify each DG rejects a load greater than or equal to its associated single largest post-accident load, and: a. Following load rejection, the frequency is < 63 Hz;b. Within 3 seconds following load rejection, the voltage is > 3740 V and < 4580 V; and c. Within 3 seconds following load rejection, the frequency is > 58.8 Hz and < 61.2 Hz.----------
i SR 3.8.1.10 Verify each DG does not trip and generator speed is maintained
< 500 rpm during and following a load rejection of > 5600 kW and < 5750 kW.(continued)
Amendment Nos 13 5 Catawba Units 1 and 2 3.8.1-8 AC Sources -Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.1.11-------------------
NOTES ----------------
- 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.Verify on an actual or simulated loss of offsite power signal: a. De-energization of emergency buses;b. Load shedding from emergency buses;c. DG auto-starts from standby condition and: 1. energizes the emergency bus in< 11 seconds, 2. energizes auto-connected shutdown loads through automatic load sequencer, 3. maintains steady state voltage> 3740 V and < 4580 V, 4. maintains steady state frequency> 58.8 Hz and < 61.2 Hz, and 5. supplies auto-connected shutdown loads for > 5 minutes.(continued)
Catawba Units 1 and 2 3.8.1-9 Amendment Nos 1 AC Sources -Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.1.12.............------...........------
NOTE ---- --..... -----------------
All DG starts may be preceded by prelube period.Verify on an actual or simulated Engineered Safety Feature (ESF) actuation signal each DG auto-starts from standby condition and: a. In < 11 seconds after auto-start and during tests, achieves voltage > 3740 V and < 4580 V;b. In < 11 seconds after auto-start and during tests, achieves frequency
> 58.8 Hz and < 61.2 Hz;c. Operates for > _5 minutes; and d. The emergency bus remains energized from the Offsite power system.18----------------(continued)
Amendment Catawba Units 1 and 2 3.8.1-10 AC Sources -Operating 3.8.1-I'IU : r UIMIIVI- I s .ýconuinueoj SURVEILLANCE FREQUENCY SR 3.8.1.13 Verify each DG's non-emergency automatic trips are bypassed on actual or simulated loss of voltage signal on the emergency bus concurrent with an actual or simulated ESF actuation signal.SR 3.8.1.14---------------------------------
NOTE -------------------------------
Momentary transients outside the load and power factor ranges do not invalidate this test.Verify each DG operating at a power factor < 0.9 operates for > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> loaded > 5600 kW and<'5750 kW.6118j o4 (continued)
Catawba Units 1 and 2 3.8.1-11 Amendment Nos. (236 /232 AC Sources -Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY+SR 3.8.1.15---NOTES---1. This Surveillance shall be performed within 5 minutes of shutting down the DG after the DG has operated > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> loaded > 5600 kW and< 5750 kW or until operating temperature is stabilized.
Momentary transients outside of load range do not invalidate this test.2. All DG starts may be preceded by an engine prelube period.Verify each DG starts and achieves, in < 11 seconds, voltage > 3740 V, and frequency
> 57 Hz and maintains steady state voltage > 3740 V and < 4580 V and frequency
> 58.8 Hz and < 61.2 Hz.SR 3.8.1.16 This Surveillance shall not be performed in MODE 1, 2, 3, or 4.Verify each DG: a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power;b. Transfers loads to offsite power source; and c. Returns to standby operation.(continued)
Catawba Units 1 and 2 3.8.1-12 AmnmetNos.E AC Sources -Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.1.17-----------------
NOTE This Surveillance shall not be performed in MODE 1, 2, 3, or4.Verify, with a DG operating in test mode and connected to its bus, an actual or simulated ESF actuation signal overrides the test mode by: a. Returning DG to standby operation; and b. Automatically energizing the emergency load from offsite power.SR 3.8.1.18 Verify interval between each sequenced load block is within the design interval for each automatic load sequencer.(continued)
Catawba Units 1 and 2 3.8.1-13 Amendment Nos(ý AC Sources -Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.1.19-----------------
NOTES---------
- 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.Verify on an actual or simulated loss of offsite power signal in conjunction with an actual or simulated ESF actuation signal: a. De-energization of emergency buses;b. Load shedding from emergency buses; and c. DG auto-starts from standby condition and: 1. energizes the emergency bus in< 11 seconds, 2. energizes auto-connected emergency loads through load sequencer, 3. achieves steady state voltage > 3740 V and < 4580 V, 4. achieves steady state frequency
> 58.8 Hz and < 61.2 Hz, and 5. supplies auto-connected emergency loads for > 5 minutes.(continued)
Catawba Units 1 and 2 3.8.1-14 Amendment Nos( R AC Sources-'
Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY ,SR 3.8.1.20-NOTE --- ------------
All DG starts may be preceded by an engine prelube period.Verify when started simultaneously from standby condition, each DG achieves, in < 11 seconds, voltage of> 3740 V and frequency of.> 57 Hz and maintains steady state voltage > 3740 V and < 4580 V, and frequency> 58.8 Hz and < 61.2 Hz.5Ei- k.S __ -ýý_I Catawba Units 1 and 2 3.8.1-15 Amendment Nos.
Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. One or more DGs with D.1 Restore stored fuel oil 30 days new fuel oil properties properties to within limits.not within limits.E. One or more DGs with E.1 Restore starting air 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> starting air receiver' receiver pressure to pressure < 210 psig and > 210 psig.> 150 psig.F. Required Action and F.1 Declare associated DG Immediately associated Completion inoperable.
Time not met.OR One or more DGs diesel fuel oil, lube oil, or starting air subsystem not within limits for reasons other than Condition A, B, C, D, or E.SURVEILLANCE REQUIREMENTS, SURVEILLANCE JFREQUENCY SR 3.8.3.1 Verify the fuel oil storage system contains > 77,7100 gal of fuel for each DG.(continued)
Catawba Units 1 and 2 3.8.3-2 Amendment Nos(ý;ý Diesel Fuel Oil, Lube Oil,' and Starting Air 3.8.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.3.2 Verify lubricating oil inventory is > 400 gal.W q fs! :- ý4: .ýSR 3.8.3.3 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 SR 3.8.3.4 Verify each DG air start receiver pressure is > 210 psig.SR 3.8.3.5 Check for and remove accumulated water from each fuel oil storage tank.Catawba Units 1 and 2 3.8.3-3 Amendment Nos.E 51 DC Sources -Operating 3.8.4 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. A and/or D channel of D.1 Enter applicable Immediately DC electrical power Condition(s) and Required subsystem inoperable.
Action(s) of LCO 3.8.9,"Distribution Systems-AND Operating", for the associated train of DC Associated train of DG electrical power distribution DC electrical power subsystem made subsystem inoperable, inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify DC channel and DG battery terminal voltage is> 125 V on float charge.1N5E1?7)S R 3.8.4.2 Not used.SR 3.8.4.3 Verify no visible corrosion at the DC channel and DG battery terminals and connectors.
(: IAl5FJ~T~OR Verify battery connection resistance of these items is< 1.5 E-4 ohm.(continued)
Amendment Nos.Catawba Units 1 and 2 3.8.4-2 DC'Sources
-Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.8.4.4 Verify DC channel and DG battery cells, cell plates, and 18 racks show no visual indication of physical damage or abnormal deterioration that could degrade battery performance.
SR 3.8.4.5 Remove visible terminal corrosion, verify DC channel and 18 on DG battery. cell to cell and terminal connections are clean and tight, and are coated with anti-corrosion material.SR 3.8.4.6 Verify DC channel and DG battery connection resistance 18 0'8-is < 1.5 E-4 ohm.SR 3.8.4.7 Verify each DC channel battery charger supplies 9)s> 200 amps and the DG battery charger supplies > 75 amps with each charger at > 125 V for > 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.SR 3.8.4.8 --------------------
NOTES ----------------
- 1. The modified performance discharge test in SR 3.8.4.9 may be performed in lieu of the service test in SR 3.8.4.8.2. This Surveillance shall not be performed for the DG batteries in MODE 1, 2, 3, or 4.Verify DC channel and DG battery capacity 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.(continued)
Catawba Units 1 and 2 3.8.4-3 Amendment No DC Sources -Operating 3.8.4 gIIP\/FII I AM('F PFC~IJIRFMFNJT$~ (r~nntinti~rI~
SURVEILLANCE FREQUENCY SR 3.8.4.9-----------
7-------------
NOTE ---------------
This Surveillance shall not be performed for the DG batteries in MODE 1, 2, 3, or 4.Verify DC channel and DG battery capacity is > 80% of the manufacturer's rating when subjected to a performance discharge test or a modified performance discharge test.AND 18 months when battery shows degradation or has reached 85% of expected life with capacity < 100%of manufacturer's rating AND..--------
NOTE --------Not applicable to DG batteries 24 months when battery has reached 85% of the expected life with capacity.>
100% of manufacturer's rating Catawba Units 1 and 2 3.8.4-4 Amendment Nos.8 Batte'y Cell Parameters 3.8.6 t... IF,~ .'r-mI I A LIrSr rr-tI IIr~r-1. ar-I. I-I-t-~0uIr¶VrILL/MINL F I'r- Ur--IVlr'l I , SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters of the channels of DC and -DG batteries meet Table 3.8.6-1 Category A limits.SR 3.8.6.2 Not used.SR 3.8.6.3 Verify battery cell parameters of the channels of DC and DG batteries meet Table 3.8.6-1 Category B limits.AND Once within 7 days after a battery discharge<110 V AND Once within 7 days after a battery overcharge
> 150 V-I SR 3.8.6.4 Verify average electrolyte temperature for the channels of DC and DG batteries of representative cells is > 60 0 F.Catawba Units 1 and 2 3.8.6-4 Amendment Nos e I10 CHANGES THIS PAGE."FOR INFORMATION ONLY Battery Cell Parameters 3.8.6 Table 3.8.6-1 (page 1 of 1)Battery Cell Parameters Requirements CATEGORY A: CATEGORY C: LIMITS FOR EACH CATEGORY B: ALLOWABLE DESIGNATED LIMITS FOR EACH LIMITS FOR EACH PARAMETER PILOT CELL CONNECTED CELL CONNECTED CELL Electrolyte Level > Minimum level > Minimum level Above top of plates, indication mark, and indication mark, and and not overflowing
<Y4 inch above < 1/4 inch above maximum level maximum level indication mark(a) indication mark(a)Float Voltage > 2.13 V > 2.13 V > 2.07 V Specific Gravity(b)(c)
> 1.200 > 1.195 Not more than 0.020 below average of all AND connected cells or> 1.195 Average of all connected cells AND> 1.205 Average of all connected cells> 1.195 (a) It is acceptable for the electrolyte level to temporarily increase above the specified maximum during equalizing charges provided it is not overflowing.(b) Corrected for electrolyte temperature and level. Level correction is not required, however, when battery charging is < 2 amps when on float charge.(c) A battery charging current of < 2 amps when on float charge is acceptable for meeting specific gravity limits following a battery recharge, for a maximum of 7 days. When charging current is used to satisfy specific gravity requirements, specific gravity of each connected cell shall be measured prior to expiration of the 7 day allowance.
Catawba Units 1 and 2 3.8.6-5 Amendment Nos. 223/218 Inverters
-Operating 3.8.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct inverter voltage and alignment to required AC vital buses.Catawba Units 1 and 2 3.8.7-2 Amendment Noo"'17ý/4165 Inverters
-Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued)
A.2.3 Suspend operations Immediately involving positive reactivity additions that could result in loss of required SDM or required boron concentration.
AND A.2.4 Initiate action to restore Immediately required inverters to OPERABLE status.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct voltage and alignment to required AC vital bus.UD, 4ý9sýCatawba Units 1 and 2 3.8.8-2 Amendment Nos. Eý )
Distribution Systems -Operating 3.8.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to required AC, DC channel, DC train, and AC vital bus electrical power distribution subsystems.
9ý2 Catawba Units. 1 and 2 3.8.9-3 Amendment No-Distribution Systems -Shutdown 3.8.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued)
A.2.4 Initiate actions to restore Immediately required AC, channels of DC, DC trains, and AC vital bus electrical power distribution subsystems to OPERABLE status.AND A.2.5 Declare associated Immediately required residual heat removal subsystem(s) inoperable and not in operation.
AND A.2.6 Declare affected Low Immediately Temperature Overpressure Protection feature(s) inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to required AC, DC channel, DC train, and AC vital bus electrical power distribution subsystems.
G3 ýJ ___________________
Catawba Units 1 and 2 3.8.10-2 Amendment No1165 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 rofueling canal, and the refueling cavity shall be maintained within the limit I : fied in the COLR.--...-.-.-
..-------------------
NOTE-------
Only applicable to the refueling canal and refueling cavity when connected to the RCS.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.Catawba Units 1 and 2 3.9.1-1 Amendment Nos.
Nuclear Instrumentation 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4-SR 3.9.2.1 Perform CHANNEL CHECK.SR 3.9.2.2-NO T E ---------------------------------
Neutron detectors are excluded from CHANNEL CALIBRATION.
ON .ýdrxtl-Perform CHANNEL CALIBRATION.
Catawba Units 1 and 2 3.9.2-2 Amendment Nos. (1 257ý20, Containment Penetrations 3.9.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. One or more CPES B.1 Restore CPES train(s) 7 days train(s) heater heater to OPERABLE inoperable, status.OR B.2 Initiate action in 7 days accordance with Specification 5.6.6.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify each required containment penetration is in the required status.SR 3.9.3.2 Operate each CPES for > 10 continuous hours with the heaters operating.
SR 3.9.3.3 Perform required CPES filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP). the VFTP 0 Catawba Units I and 2 3.9.3-2 Amendment Nos (173//65 RHR and Coolant Circulation
-High Water Level 3.9.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued)
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 reactor coolant at a flow rate of > 1000 gpm and RCS temperature is < 140'F.(Catawba Units 1 and 2 3.9.4-2 Amendment NoU .13 5 RHR and Coolant Circulation
-Low Water Level 3.9.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued)
B.2 Initiate action to restore Immediately one RHR loop to operation.
AND B.3 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 IFREQUENCY SR 3.9.5.1 Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of > 1000 gpm and RCS temperature is < 140'F.SR 3.9.5.2 Verify correct breaker alignment and indicated power available to the required RHR pump that is not in operation.
Catawba Units 1 and 2 3.9.5-2 Amendment Nos. 0173/!ý:)
Refueling Cavity Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Refueling Cavity Water Level LCO 3.9.6 APPLICABILITY:
Refueling cavity water level shall be maintained
> 23 ft above the top of reactor vessel flange.During CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, During movement of irradiated fuel assemblies within containment.
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 top of reactor vessel flange.K= -Catawba Units 1 and 2 3.9.6-1 Amendment Unborated Water Source Isolation Valves 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Unborated Water Source Isolation Valves LCO 3.9.7 Each valve used to isolate unborated water sources shall be secured in the closed position.APPLICABILITY:
MODE 6.ACTIONS----------------------------------------------------
I\ 1t j r -----------------------------------
---Separate Condition entry is allowed for each unborated water source isolation valve.CONDITION REQUIRED ACTION COMPLETION TIME A. --------NOTE ------------
A.1 Suspend CORE Immediately Required Action A.3 ALTERATIONS.
must be completed whenever Condition A is AND entered.A.2 Initiate actions to secure Immediately valve in closed position.One or more valves not secured in closed AND position.A.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 /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify each valve that isolates unborated water sources is secured in the closed position.31 a Catawba Units I and 2 319.7-1 Amendment Nos. C15 ,;4&
Programs and Manuals 5.5 Programs)and Manuals (continued) 5.5.16 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Area Ventilation System (CRAVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge.
The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident.
The program shall include the following elements: a. The definition of the CRE and the CRE boundary.b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
- c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197,"Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1. and C.2. of Regulatory Guide 1.197, Revision 0.d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CRAVS, operating at a makeup flow rate of < 4000 cfm, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.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, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs 3 c and d, respectively.
Catawba Units 1 and 2 5.5-15 Am-endment ATTACHMENT 4 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES INSERTS INSERT2 The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.REVIEWER'S NOTE: Text deleted and replaced by Insert 2 will be relocated to the Surveillance Frequency Control Program (SFCP) document(s) per TSTF-425.Thus, there are instances in these mark-ups where deleted text is edited for future use in the SFCP. The words "For SFCP addition only" will accompany inserted text that will be relocated to the SFCP. This inserted text will be cross-hatched to indicate it is not to be inserted on the Bases page.
SDM B 3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)
In MODE 2 with kef < 1.0 and MODES 3, 4, and 5, SDM is verified by performing a reactivity balance calculation, considering the listed reactivity effects: a. RCS boron concentration;
- b. Control bank position;c. RCS average temperature;
- d. Fuel burnup based on gross thermal energy generation;
- e. Xenon concentration;
- f. Samarium concentration; and g. Isothermal temperature coefficient (ITC).Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.hFre enc of 24 ours' base on the enera slo cha e in r quire bor con ntrati n an he low robabacent occu ing thout e re ired M. T sallo tim for t oper or to.co1 ct t requi ed da ,wi inclu s pe rmin *a boon 'nce ration nalysj , an_ ompi e the alcul on.REFERENCES
- 1. 10 CFR 50, Appendix A, GDC 26.2. UFSAR, Section 15.1.5.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 4. UFSAR, Section 15.4.6.5. 10 CFR 50.67.Catawba Units 1 and 2 B 3.1.1-6 Revision No.(R Core Reactivity B 3.1.2 BASES ACTIONS (continued)
B.1 If the core reactivity cannot be restored to within the 1 % Ak/k limit, 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 />. If the SDM for MODE 3 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 3 from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE SR 3.1.2.1 REQUIREMENTS 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.
The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization of predicted core reactivity to the measured value must take place within the first 60 effective full power days (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 __esi n calculat s. Te re ire ub e ue Freq ncy 31 FP isac pt e, sed nct sl cha ges ue t fuel eple i n n thpre nc f o in~(T AFD etc. forp mp ndi itio f a no al REFERENCES
- 1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specification, (c)(2)(ii).
Catawba Units 1 and 2 B 3.1.2-5 Revision No-@
Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued)
D.1.1 and D.1.2 More than one control rod becoming misaligned from its group average position is not expected, and has the potential to reduce SDM.Therefore, SDM must be evaluated.
One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity, as described in the Bases or LCO 3.1.1. The required Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action.This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SDM is restored.D.2 If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions.
The unit must be brought to a MODE or Condition in which the LCO requirements are not applicable.
To achieve this status, the unit 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 />.The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE SR 3.1.4.1 REQUIREMENTS
_________[Verific ion that in ividua rod po tions re w in alig ent limi at a Freq ncy of 12 ours ovides/ hist ry th allows he oper or to/ det t a rod th is be nning t, 'devi le fro its ex ected p sition. If th ro position d viatior onitor is mo erab a Fr uency 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> complish the s e go ,The'spec ied Fr uenc akes into ccount ot r rod osition for tion at is ontinu sly avail le to e operator i the c trol r m, s tht uring ctual motio devi ions can im diatel e de cted .SR 3.1.4.2 .Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2, tripping each control rod would result in radial or axial power tilts, or oscillations.
Exercising each Catawba Units 1 and 2 B 3.1.4-7 Revision No.
Rod Group Alignment Limits B 3.1.4 BASES SURVEILLANCE REQUIREMENTS (continued) individual control rodas92 srovides increased confidence that all rods continue to be OPABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each control rod by 1 0 steps will got cause radial or axial power tilts or oscillations, to occur.,,9t cskad raetio othe ifor atio-t op,, tor.. t .c .om ..dS. 3.1 .1, i- I R LIT of th, rods Between required p ormances of SR 3.1.4.2 (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. This may be by verification of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism.
During performance of the Control Rod Movement periodic test, there have been some "Control Malfunctions" that prohibited a control rod bank or group from moving when selected, as evidenced by the demand counters and DRPI. In all cases, when the control malfunctions were corrected, the rods moved freely (no excessive friction or mechanical interference) and were trippable.
SR 3.1.4.3 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.
Since a removal of the reactor vessel head has the potential to change component alignments affecting rod drop times, measuring drop times prior to the next criticality following any such 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 all RCPs operating and the average moderator temperature
> 551'F 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.Catawba Units 1 and 2 B 3.1.4-8 Revision No.-
Shutdown Bank Insertion Limits B 3.1.5 BASES ACTIONS (continued)
B. 1 If the shutdown banks cannot be restored to within their insertion limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the unit must be brought to a MODE where the LCO is not applicable.
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 Verification that the shutdown banks are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown banks will be available to shut down the reactor, and the required SDM will be maintained following a reactor-trip.
This SR and Frequency ensure that the shutdown banks are withdrawn before the control banks are withdrawn during a unit startup.REFERENCES
- 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, and GDC 28.2. 10 CFR 50.46.3. UFSAR, Section 15.4.4. 10 CFR 50.36, Technical Specification, (c)(2)(ii).
Catawba Units 1 and 2 B 3.1.5-4 Revision NoOl Control Bank Insertion Limits B 3.1.6 BASES ACTIONS (continued) required MODE from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE REQUIREMENTS SR 3.1.6.1 This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits.The estimated critical position (ECP) depends upon a number of factors, one of which is xenon concentration.
If the ECP was calculated long before criticality, xenon concentration could change to make the ECP substantially in error. Conversely, determining the ECP immediately before criticality could be an unnecessary burden. There are a number of unit parameters requiring operator attention at that point. Verifying the ECP calculation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to criticality avoids a large error from changes in xenon concentration, but allows the operator some flexibility to schedule the ECP calculation with other startup activities.
SR 3.1.6.2 nor aIly, ýeryIt-tle
/odliotio6 oc_ rs0i 12 ý6urs. Wthe insertion limit mon-itor-bcomnes ino"perable, verification of the control bank position at a Frequency of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is sufficient to detect control banks that may be approaching the insertion limits.SR 3.1.6.3 When control banks are maintained within their insertion limits as checked by SR 3.1.6.2 above, it is unlikely that their sequence and overlap will not be in accordanc with re uir s roviin the COLR. ,ncv oj-1 hfrs. __sist.rwitl e Catawba Units 1 and 2 B 3.1.6-5 Revision.
No-9 PHYSICS TESTS Exceptions B 3.1.8 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.1.8.2 Verification that the RCS lowest loop Tavg is > 541OF will ensure that the unit is not operating in a condition that could invalidate the safetY SR 3.1.8.3 Verification that THERMAL POWER is < 5% RTP will ensure that the plant is not o eratin ji e co A SR 3.1.8.4 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects: a. RCS boron concentration;
- b. Control bank position;c. RCS average temperature;
- d. Fuel burnup based on gross thermal energy generation;
- e. Xenon concentration;
- f. Samarium concentration; and g. Isothermal temperature coefficient (ITC).Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.Catawba Units 1 and 2 B 3.1.8-5 Revision No 12 FQ(X,Y,Z)B 3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued) verification.
It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FQ was last measured.SR 3.2.1.1 Verification that FMQ(X,Y,Z) is within its specified steady state limits involves either increasing FMQ(X,Y,Z) to allow for manufacturing tolerance, K(BU), and measurement uncertainties for the case where these factors are not included in the FQ limit. For the case where these factors are included, a direct comparison of FMo(X,Y,Z) to the F 0 limit can be performed.
Specifically, FMQ(X,Y,Z) is the measured value of Fo(X,Y,Z) obtained from incore flux map results. Values for the manufacturing tolerance, K(BU), and measurement uncertainty are specified in the COLR.The limit with which FMQ(X,Y,Z) is compared varies inversely with power above 50% RTP and directly with functions called K(Z) and K(BU)provided in the COLR.If THERMAL POWER has been increased by > 10% RTP since the last determination of FMa(X,Y,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 FMQ(X,Y,Z) values have decreased sufficiently with power increase to stay within. the LCO limits).fhe Fre 'ency of 1 'EFPID rs'eut 1moito thi ichange of pocr Fdistrib ion with/re burn tbecause uch cha'n are'slow well '
whe .the pla ;Jso oera~t~ in acor ~fce with t SR 3.2.1.2 and 3.2.1.3 The nuclear design process includes calculations performed to determine that the core can be operated within the F 0 (X,Y,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, is determined by a maneuvering analysis (Ref. 5).Catawba Units 1 and 2 B 3.2.1-9 Revision No(D' NO CHANGES THIS PAGE. F(X,Y,Z)FOR INFORMATION ONLY B 3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The limit with which FMo(X,Y,Z) is compared varies and is provided in the COLR. No additional uncertainties are applied to the measured FQ(X,Y,Z) because the limits already include uncertainties.
FLQ(X,Y,Z)° and FLQ(X,Y,Z)RPs limits are not applicable for the following axial core regions, measured in percent of core height: a. Lower core region, from 0 to 15% inclusive; and b. Upper core region, from 85 to 100% inclusive.
The top and bottom 15% 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 FMo(X,Y,Z) is evaluated and found to be within the applicable transient limit, an evaluation is required to account for any increase to FMQ(X,Y,Z) that may occur and cause the FQ(X,Y,Z) limit to be exceeded before the next required FQ(X,Y,Z)evaluation.
In addition to ensuring via surveillance that the heat flux hot channel factor is within its limits when a measurement is taken, there are also requirements to extrapolate trends in both the measured hot channel factor and in its operational and RPS limits. Two extrapolations are performed for each of these two limits: 1. The first extrapolation determines whether the measured heat flux hot channel factor is likely to exceed its limit prior to the next performance of the SR.2. The second extrapolation determines whether, prior to the next performance of the SR, the ratio of the measured heat flux hot channel factor to the limit is likely to decrease below the value of that ratio when the measurement was taken.Each of these extrapolations is applied separately to each of the operational and RPS heat flux hot channel factor limits. If both of the extrapolations for a given limit are unfavorable, i.e., if the extrapolated factor is expected to exceed the extrapolated limit and the extrapolated factor is expected to become a larger fraction of the extrapolated limit Catawba Units 1 and 2 B 3.2.1-10 Revision No. 0 FQ(X,Y,Z)B 3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued) than the measured factor is of the current limit, additional actions must be taken. These actions are to meet the FQ(X,Y,Z) limit with the last FMQ(X,Y,Z) increased by the appropriate factor specified in the COLR or to evaluate F 0 (X,Y,Z) prior to the projected point in time when the extrapolated Values are expected to exceed the extrapolated limits.These alternative requirements attempt to prevent FQ(XY,Z) from exceeding its limit for any significant period of time without detection using the best available data. FMQ(XY,Z) is not required to be extrapolated for the initial flux map taken after reaching equilibrium.
conditions since the initial flux map establishes the baseline measurement for future trending.
Also, extrapolation of F M Q(X,Y,Z) limits are not valid for core locations that were previously rodded, or for core locations that were previously within +/-2% of the core height about the demand position of the rod tip.FQ(X,YZ) i's 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(X,Y,Z) is within its limit at higher power levels.The Survei ance Frequency/f 31 EFPD is to rrnitor the hange 1 power distributi5 with core bu up. The Su ilance may done. 0 frequentlye hnuiredal Spfsultsiofns, Fc Z)evaluati The Feque~ncy ofl 3/FPD is adeq te to mnitor he change ,2power 5dstriutinbecause such a1chan1 As cientl low, whens plantis eMOpe tedinaccor dg forice Optherain ts of adWestigoingsfactors b 31 day Yu rveilla ii "-.
.11. 10 CFR 50.46.2- UFSAR Section".15.4.8.
- 3. 10 CFR 50, Appendix A, GDC 26.4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 5. DPC-NE-201 1PA "Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse.
Reactors".
Catawba Units 1 and 2 B 3.2.1-11 Revision No. "
FH(X.Y)B 3.2.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.2.2.1 The value of FAH(X,Y) is determined by using the movable incore detector system to obtain a flux distribution map at any THERMAL POWER greater than 5% RTP. A computer program is used to process the measured 3-D power distribution to calculate the steady state FLj(X,Y)"c limit which is compared against FAHK(X,Y).
FM,(X,Y) 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 FMm(X,Y) is within its limit at high power levels.The 31 FPD Fre ency is acce able because* e power stribution'" cha es relativel slowly-over t s amount of f I burnup Accordin th* Frequency
.short enoug that the FAH(.) limit nn e ceeded for ny significan eriod of operp-i-n.
SR 3.2.2.2 The nuclear design process includes calculations performed to determine that the core can be operated within the FýH(X,Y) 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 is a limit called FLAH (X,Y)sv. This Surveillance compares the measured to the Surveillance limit to ensure that safety analysis limits are maintained.
This Surveillance has been modified by a Note that may require that more frequent surveillances be performed.
If FMm(X,Y) is evaluated and found to be within its surveillance limit, an evaluation is required to account for any increase to FM (X,Y) that may occur and cause the FM(X,Y)suRv limit to be exceeded before the next required FH(X,Y)SuRv evaluation.
In addition to ensuring via surveillance that the enthalpy rise hot channel factor is within its steady state and surveillance limits when a measurement is taken, there are also requirements to extrapolate trends in both the measured hot channel factor and in its surveillance limit. Two extrapolations are performed for this limit: Catawba Units 1 and 2 B 3.2.2-8 Revision NoZ,ý FAH(X,Y)B 3.2.2 BASES SURVEILLANCE REQUIREMENTS (continued)
- 1. The first extrapolation determines whether the measured enthalpy rise hot channel factor is likely to exceed its surveillance limit prior to the next performance of the SR.2. The second extrapolation determines whether, prior to the next performance of the SR, the ratio of the measured enthalpy rise hot channel factor to the surveillance limit is likely to decrease below the value of that ratio when the measurement was taken.Each of these extrapolations is applied separately to the enthalpy rise hot channel factor surveillance limit. If both of the extrapolations are unfavorable, i.e., if the extrapolated factor is expected to exceed the extrapolated limit and the extrapolated factor is expected to become a larger fraction of the extrapolated limit than the measured factor is of the current limit, additional actions must be taken. These actions are to meet the FMAH(X,Y) limit with the last FMAH(X,Y) increased by a factor of 1.02, or to evaluate FMAH(X,Y) prior to the point in time when the extrapolated values are expected to exceed the extrapolated limits. These alternative requirements attempt to prevent FMAH(X,Y) from exceeding its limit for any significant period of time without detection using the best available data.F MAH(X,Y) is not required to be extrapolated for the initial flux map taken after reaching equilibrium conditions since the initial flux map establishes the baseline measurement for future trending.FMAH(X,Y) 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 FMAH(X,Y) is within its limit at high power levels.The Surve)f'ance Freq ncy of 31 E D is adeq te to monitor change 9 power distriution with c/re burnup. 7he Surveillanc may be done m re freauentl4 if reauired bv the results f FMAW(X.Y) eva uations.REFERENCES
- 1. UFSAR Section 15.4.8 2. 10 CFR 50, Appendix A, GDC 26.Catawba Units 1 and 2 B 3.2.2-9 Revision Noo NO CHANGES THIS PAGE.FOR INFORMATION ONLY AFD B 3.2.3 BASES LCO (continued)
Signals are available to the operator from the Nuclear Instrumentation System (NIS) excore neutron detectors (Ref. 3). Separate signals are taken from the top and bottom detectors.
The AFD is defined as the difference in normalized flux signals between the top and bottom excore detectors in each detector well. For convenience, this flux difference is converted to provide flux difference units expressed as a percentage and labeled as %A flux or %AI.The AFD limits are provided in the COLR. The AFD limits do not depend on the target flux difference.
However, the target flux difference may be used to minimize changes in the axial power distribution.
Violating this LCO on the AFD could produce unacceptable consequences if a Condition 2, 3, or 4 event occurs while the AFD is outside its specified limits.APPLICABILITY The AFD requirements are applicable in MODE 1 greater than or equal 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 maneuvering analysis 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 condition for which the value of the AFD is not important in the applicable safety analyses.
A Completion Time of 30 minutes is reasonable, based on operating experience, to reach 50% RTP without challenging plant systems.SURVEILLANCE SR 3.2.3.1 REQUIREMENTS The AFD is monitored on an automatic basis using the unit process computer, which has an AFD monitor alarm. The computer determines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels is outside its specified limits.Catawba Units 1 and 2 B 3.2.3-3 Revision No. 0 AFD B 3.2.3 BASES SURVEILLANCE REQUIREMENTS (continued) l-his Surveillance verifies that the AFD, as indicated by the NIS excore channel, is within its specified limits and is consistent with the status of the AFD monitor alarm. With the AFD monitor alarm inoperable, the AFD is monitored every hour to detect operation outside its limit. The Frequency of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is based on operating experience regarding the amount of time reaqired to vary the AFD, and the fact that the FD is closely monitored.
-,-.Nith the AFD monitor alarm OPERABLE,the urveian e: Frequency ofV days is adi'uate c sidering-a e AFD is monitor d by a comptt( ard any de/ation fr requireme(fs is REFERENCES
- 1. DPC-NE-201 1 PA, "Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors".
- 2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 3. UFSAR, Chapter 7.Catawba Units 1 and 2 B 3.2.3-4 Revision No.0 QPTR B 3.2.4 BASES ACTIONS (continued) reaching RTP. As an added precaution, if the core power does not reach RTP within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, but is increased 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 /> of the time when the more restrictive of the power level limit determined by Required Action A.1 or A.2 is exceeded.
These Completion Times are intended to allow adequate time to increase THERMAL POWER to above the more restrictive limit of Required Action A.1 or A.2, while not permitting the core to remain with unconfirmed power distributions for extended periods of time.Required Action A.7 is modified by a Note that states that the peaking factor surveillances must be done after the excore detectors have been calibrated to show zero tilt (i.e., Required Action A.6). 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 calibrated to show zero tilt and the core returned to power.B.1 If Required Actions A.1 through A.7 are not completed within their associated Completion Times, the unit 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 SR 3.2.4.1 is modified by three Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER-is <75% RTP and the input from one Power Range Neutron Flux channel is inoperable.
Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1. Note 3 states that the 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 exceeding 50% RTP. This is necessary to establish core conditions necessary to provide meaningful calculation.
This Surveillance verifies that the QPTR, as indicated by the Nuclear, Instrumentation System (NIS) excore channels, is within its limits. e Catawba Units 1 and 2 B 3.2.4-5 Revision No-&
QPTR B 3.2.4 SURVEILLANCE REQUIREMENTS (continued)
When the QPTR alarm is inoperable, the Frequency is increased to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This Frequency is adequate to detect any relatively slow changes in QPTR, because for those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.The QPTR alarm is inoperable for the duration of excore channel calibrations performed for agreement with incore detector measurements.
SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is required only when the input from one or more Power Range Neutron Flux channels are inoperable and the THERMAL POWER is >_ 75% RTP.With an 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 capabi er tilts in some quadrants i ecreased.
Perform~ingZ 324,aeýFreucy o urs provid an accurate ernative m~ans for entn than tilt rmi s within its IiZs.For purposes of monitoring the QPTR when one power range c annel is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR and any previous data indicating a tilt. The incore detector monitoring is performed with a full incore flux map or two sets of four thimble locations with quarter core symmetry.
The two sets of four symmetric thimbles is a set of eight unique detector locations.
These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.The symmetric thimble flux map can be used to generate symmetric thimble "tilt." This can be compared to a reference symmetric thimble tilt, from the most recent full core flux map, to generate an incore tilt.Therefore, incore tilt can be used to confirm that QPTR is within limits.With one or more NIS channel inputs to QPTR inoperable, the indicated tilt may be changed from the value indicated with all four channels OPERABLE.
To confirm that no change in tilt has actually occurred, Catawba Units 1 and 2 B 3.2.4-6 Revision No[D NO CHANGES THIS PAGE.FOR INFORMATION ONLY QPTR~B 3.2.4 BASES i_ __.....SURVEILLANCE REQUIREMENTS (continued) which might cause the QPTR limit to be exceeded, the incore result may be compared against previous flux maps either using the symmetric thimbles as described above or a complete flux map. Nominally, quadrant tilt from the Surveillance should be within 2% of the tilt shown by the most recent flux map data.REFERENCES
- 1. 10 CFR 50.46.2. UFSAR Section 15.4.8.3. 10 CFR 50, Appendix A, GDC 26.4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.2.4-7 Revision No. 1 RTS Instrumentation B 3.3.1 BASES ACTIONS (continued)
U.1 With two RTS trains inoperable, no automatic capability is available to shut down the reactor, and immediate plant shutdown in accordance with LCO 3.0.3 is required.SURVEILLANCE The SRs for each RTS Function are identified by the SRs column of REQUIREMENTS Table 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 I, Train A and Train B must be examined.Similarly, Train A and Train B must be examined when testing Channel II, Channel III, and Channel IV (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.
Performing the Neutron Flux Instrumentation surveillances meets .the License Renewal Commitments for License Renewal Program for High-Range Radiation and Neutron Flux Instrumentation Circuits per UFSAR Chapter 18, Table 18-1 and License Renewal Commitments specification QNS-1274.00-00-0016.
SR 3.3.1.1, Performance of the CHANNEL CHEC re 12 urI'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 two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff based on a combination of the channel instrument uncertainties, including indication Catawba Units 1 and 2 B 3.3.1-42 Revision No 0 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) 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. gi r 0 AA d, tos, Th red sed plra i"g e ori-ncl that de
]channl failure is rare. 5e CHANNIE CHECK suppl ents, less f mal,, ibut frequent, ch ks of chan drng o operationa se the isPlays associ ed with r equie cne SR 3.3.1.2 SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS channel output( 2 os 7 If the calorimetric exceeds the NIS channel'output by >- % , the NIS is not declared inoperable, but must be adjusted.
If the NIS channel output cannot be properly adjusted, the channel is declared inoperable.
Two Notes modify SR 3.3.1.2. The first Note indicates that the NIS channel output shall be adjusted consistent with the calorimetric results if the absolute difference between the NIS channel output and the calorimetric is > 2% RTP. The second Note clarifies that this Surveillance is required only if reactor power is >_ 15% RTP and that 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is allowed' for completing the first Surveillance after reaching 15% RTP. At lower power levels, calorimetric data are inaccurate.
The Frequen of every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is dequate. It is based on'operating perience, considerin instrument reliability an perating history ta for instrument drif. Together these factor emonstrate the cha e in the absolute diffe Iice between NIS and eat balance: c culated powers rarely xceeds 2% in any 24, h r period. Maint ing e 2% agreement is ly applicable during e ibrium condition In addition, cont room operators perio ally monitor red dant indications an alarms to detect deviati ns in channel o SR 3.3.1.3 SR 3.3.1.3 compares the-i6"66'Sys'ten'em1o the NIS channel outputýýIf the absolute difference is > 3%, the NIS channel is still OPERABLE, but must be readjusted.
Catawba Units 1 and 2 B 3.3.1-43 Revision No.t RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
If the NIS channel cannot be properly readjusted, the channel is declared inoperable.
This Surveillance is performed to verify the f(AI) input to the overtemperature AT Function and overpower AT Function.Two Notes modify SR 3.3.1.3. Note 1 indicates that the excore NIS channel shall be adjusted if the, absolute difference between the incore and excore AFD i's _> 3%. Note 2 clarifies that the Surveillance is required only if reactor power is >_ 15% RTP and that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed for completing the first Surveillance after reaching 15% RTP.TherF uen cy of. e ry 31 EFPD isa qte. It is ba d on u-nit " So'p atting experie/ ~e, considering i trument reliabi i and operatingdata for ' strument drift. rso, the slow c 3a~ges in neutron fux during the f I cycle can be d cted during th' ntevl SR 3.3.1.4 SR 3.3.1.4 is the rmance of a TADOTv AGEREDEST SI This test shall verify OPERABILITY by actuation of the end devices.The RTB test shall include separate verification of the undervoltage and shunt trip mechanisms.
Independent verification of RTB undervoltage and shunt trip Function is not required for the bypass breakers.
No capability is provided for performing such a test at power. The independent test for bypass breakers is included in SR 3.3.1.14.
The bypass breaker test shall include a local shunt trip. A Note has been added to indicate that this test must be performed on the bypass breaker prior to placing it in service.The _____ o 6 n STAG ED TES ASIS is Jdstifi ie R SR 3.3.1.5.WA~ ,.,
zoo IFICP AQ I*$ as% G SR 3.3.1.5 is the performance of an ACTUATIO GI TET. The SSPS is testeev 2ays, n a ST EST B.. ,-using the semiautomatic tester. The train being tested is placed in the bypass Catawba Units 1 and 2 B 3.3.1-44 Revision Nof RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and applicable permissives, are tested for each protection function., Freuincv o&,6verv a STAGERE-I OT BASIS'i)WcAP -15' -P- )J (4 a V'ý ov3 ror J SR 3.3.1.6 SR 3.3.1.6 is a calibration .of the excore channels to the incore channels.If the measurements do not agree, the excore channels are not declared inoperable but must be calibrated to agree with the incore detector measurements.
If the excore channels cannot be adjusted, the channels are declared inoperable.
This Surveillance is performed to verify the f(AI)input to the overtemperature AT Function and overpower AT Function.At Beginning of Cycle (BOC), the excore channels are compared to the incore detector measurements prior to exceeding 75% power. Excore detectors are adjusted as necessary.
This low power surveillance satisfies the initial performance of SR 3.3.1.6 with subsequent surveillances conducted at least every 92 EFPD.At BOC, after reaching full power steady state conditions, additional incore and excore measurements are taken at various AIconditions to determine the Mi factors. The M 1 factors are normally only determined at BOC, but they may be changed at other points in the fuel cycle if the relationship between excore and incore measurements changes significantly.
A Note modifies SR3.3.1.6.
The Note states that this Surveillance is required only if reactor power is > 75% RTP and that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed for completing the first. surveillance after reaching 75% RTP.SR 3.3.1.7 SR 3.3.1.7 is the performance of a COT<5 84,z 'as.A COT is performed on each required channel to ensure the channel will Catawba Units 1 and 2 B 3.3.1-45 R .evision No./?P, tfý/
RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) perform the intended Function.The tested portion of the loop must trip within the Allowable Values specified in Table 3.3.1-1.The setpoint shall be left set'consistent with the assumptions of the setpoint methodology.
SR 3.3.1.7 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 unit is to be in. MODE 3 with the RTBs closed for > 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> this Surveillance must be completed ia.k'hours after entry into MODE 3. "-A S R 3.311.8 SR 3.3.1.8 is the e aeof a 0 as descn ed in " 3.3.1.7, except it is modified by a Note that this test shall include verification that the P-6, during the Intermediate Range COT, and P-10, during the Power Range COT, interlocks are in their required state for the existing unit condition.
The overification is performed by visual observation of the a permissive status light in the unit control rooms The Frequec osa modified by a Note thbelo s surveillance to be satisfied if it has been performed within" 84 day, of the Frequencies prior to reactor startup and four hours after reducing power below P-10 and P-6. The Frequency of "prior to startup" ensures this surveillance is performed prior to critical operations and applies to the source, intermediate and power range low instrument channels.
The Frequency of "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" (applicable to intermediate and power range low channels) and f4 hours after reducing power below P-6" (applicable to'source range channels) allows a normal shutdown to be completed and the unit removed from the M013E of Applicability for this surveillance withouta deeyý...lhe testing required by this surveillance.
The Othereafter applies if the plant remains in the, MODE of Applicability after the initial performances of prior to reactor startup and four hours after reducing power below P-10 or P-6. The MODE of Applicability for this surveillance is < P-1 0 for the power range low and intermediate range channels and < P-6 for the source range channels.Catawba Units 1 and 2 B 3.3.1-46 Revision Noo RTS Instrumentation B 3-3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Once the unit is in MODE 3, this surveillance is no longer required.
If power is to be maintained
< P-10 or < P-6 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. Four hours is a reasonable time to complete the required testing or place the unit 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 and after reducing power into the applicable MODE (< P-10 or < P-6) for periods > 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. CTre of.aav s iu l led i e r.,,-'qý.JA'The SR is modified by a Note th t excludes verification of setpoints from the TADOT. Since this SR appli to RCP undervoltage and underfrequency relays, setpoint v fication is accomplished during the CHANNEL CALIBRATION, SR 3.3.1.10 , ýFA ý lo a.f'A-CH/ AIEL C4IAB RATION is periieed every1-months 1oýf ap -iimat every refuelin HANNEL CALIBRATION is a Snsprum nt 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 setpoint methodology.
SR 3.3.1.10 is modified by a Note stating that this test shall include verification that the time constants are adjusted to the prescribed values where applicable.
The applicable time constants are shown in Table 3.3.1-1.Catawba Units 1 and 2 B 3.3.1-47 Revision Noo 7 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.11 SR 3.3.1.11 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.10, 1 os This SR is modified by two notes. Note 1 states 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 15% RTP.The CHANNEL CALIBRATION for the source range and intermediate range neutron detectors consists of obtaining the high voltage detector plateau and discriminator .curves for source range, and the high voltage detector plateau for, intermediate range, evaluating those curves, and comparing'the curves to the manufacturer's data. Note 2 states that 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 unit must be in at least MODE 2 to perform the test for the intermodiate range detectors and MODE 1 for the power range detectors.
I.. e on /-base/on the n d to performtis urveillanc under the conditio that ap, dunn plant outage nd the potenti for an unplanned nsient if-"-e Surve ance were pe rmed with the actor at power. erating experi ce has shown ese componen usually pass the,, urveillanc-whe performed on t 18 month Fre ency SR 3.3.1.12 SR.:3.3-1.12 is the performance of a CHANNEL CALIBRATION, as described in SR 33. 1.10, 8 The F quenc s justified by assumptio~n an 18 month ibrati t i ral in l, d ot f themag de of equip t drift ine tpoint al sis. -Z4JL SR 3.3.1.13 SR 3.3.1.13 is the performance of a COT of RTS interlock<
j Th Frequ nc s ase on e own r o the interi ýims and Iticha el redund cy avai bler a has beye shown.. be ccept le throug operati exper nee Catawba Units 1 and 2 B 3.3.1-48 Revision No.,
RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.14 SR 3.3.1.14 is the performance of a TADOT of the Manual Reactor Trip and thS1 Input from ESFAS his ADO is e med e -C87.Ai. The test shall ify 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 Reactor Trip Bypass Breaker test shall include testing of the automatic undervoltage trip.Th 'uecsb edon th nownrrai;i!Iil:ýltyý,ýýiý soandth ac ptable thr h opera g ence The SR is modified by a Note that excludes verification of setpoints from 3 the TADOT. The Functions affected have no setpoints associated with them.SR 3.3.1.15 SR 3.3.1.15 is the performance of a TADOT of Turbine Trip Functions.
This TADOT is as described in SR 3.3.1.4, except that this test is performed prior to reactor startup. A Note states that this Surveillance is not required if it has been performed within the previous 31 days.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.SR 3.3.1.16 and SR 3.3,1.17 SR 3.3.1.16 and SR 3.3.1.17 verify that the individual channel/train actuation response times are less than or equal to the maximum values assumed in the accident analysis.
Response time testing acceptance criteria are included in the UFSAR (Ref. 1). Individual component response times are not modeled in the analyses.Catawba Units 1 and 2 B 3.3.1-49 Revision No.)
i RTS Instrumentation
- 'Q CX!AGES T-41S PACW.-toa ii 1 4;i. B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor to the point at which the equipment reaches the required functional state (i.e., control and shutdown rods fully inserted in the reactor core).For channels that Include dynamic transfer Functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer Function set to one, with the'resulting measured response time compared to the appropriate UFSAR response time. Aftemately, the response time test can be performed with the time constants set to their nominal value, provided the required response time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured.Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) in place, onsite, or offsite (e.g.vendor) test measurements, or (3) utilizing vendor engineering specifications.
WCAP-13632-P-A Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements" provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. in addition, while not specifically identified in the WCAP, ITT Barton 386A and 580A-0 sensors were compared to sensors which were identified.
It was concluded that the WCAP results could be applied to these two sensor types as well. Response time verification for other sensor types must be demonstrated by test.WCAP-1 4036-P-A Revision 1, ,Elimination of Periodic Protection Channel Response Time Tests provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time.The allocations for sensor, signal conditioning and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response Catawba Units 1 and 2 B 3.3.1-50 Revision No. I RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) time could be affected is replacing the sensing assembly of a transmitter.
As appropriat , each ch nels respons must be verifie every 18 months a STAG RED TEST .SIS. Testing the final actuation evices is i luded in the t ting. Testing o the RTS Ds is perform on an 18 onth frequen .Response t' es cann be deter ined during nit operation b cause equipm t operatio is requ ed to meas e response ti es. Experienc has shown hat these co ,tponents usu/ iiy pass this1 sot srveillance whe eperformed dt the.1 month Freq ency. Theref e, the Fre ue was conc -d.edtobe.
cceptable fr~t a reiblt
,_._ .. .SR 3.3.1.16 is modified by a Note stating that neutron deiedt-br6, are, excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response.
The response time of the neutron flux signal portion of the channel shall be measured from detector output or input of the first electronic component in the channel.REFERENCES
- 1. UFSAR, Chapter 7.2. UFSAR, Chapter 6.3. UFSAR, Chapter 15.4. IEEE-279-1971.
- 5. 10 CFR 50.49.6. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 7. WCAP-10271-P-A, Supplement 2, Rev. 1, June 1990.8. WCAP-13632-P-A Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements" Sep., 1995.9. WCAP-14036-P-A Revision 1, "Elimination of Periodic Protection Channel Response Time Tests" Oct., 1998.10.10 CFR 50.67.Catawba Units 1 and 2 B 3.3.1-51 Revision Noo ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE The SRs for each ESFAS Function are identified by the SRs column REQUIREMENTS of Table 3.3.2-1.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.
Note that each channel of process protection supplies both trains of the ESFAS. When testing channel I, train A and train B must be examined.Similarly, train A and train B must be examined when testing channel II, channel IIl, and channel IV (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.
SR 3.3.2.1 Performance of the CHANNEL CHECKde 72Lýensures 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 the -two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and reliability.
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.Thb eun ased e erience t-h -,'emonstrate.,/'
' [/ Ichannel fa* re is rare. ,-he CHANN 'CHECK su plements les formal,][ ~~~but mor frequent, cl;cks of chan Ils during n dal o use of/FCte i ay ss ed with theL0 require /Channels SR 3.3.2.2 SR 3.3.2.2 is the performance of an ACTUATION LOGIC TEST..)ý5 is tested every k2ýdays on a TFT Bý_`j) using.he semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the Catawba Units 1 and 2 B 3.3.2-44 Revision Ncq ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued) semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
In addition, the master relay coil is pulse tested for continuity.
This verifies that the logic modules are OPERABLE and that there is an intact voltage signal path to the master reay Fr, everyW ,92 dan'E (-a!F'G-[FED T S .. r c 1 Po 419#'/A *ALS SR 3.3.2.3 is the performance of a TADOTT.,I This test is a check of the Loss of Offsite Power Function.
Each Function is tested up to, and including, the master transfer relay coils.This test also includes trip devices that provide actuation signals directly A SFcI'to the SSPS. The SR is modified by a Note that excludes final actuation of pumps and valves to minimize plant upsets that would'occur.l-;
&6 I ("re,,qency j- a-equat ased o opera i g experie,:e co si err~ / g~ruerel=iability nd ope, an hcry dt./ SR 3.3.2.4 :\1CAT 2-SR 3.3.2.4 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity.
4GW. _ýrfo, ed y ý,c'd ton T FE B T l B _he time ,allo wed tor the testing (4 husis justified in Rteference 7 e,,41 SR 3.3.2.5 is the performance of a COT. W/ -.5P-A COT is performed on each required channel o ensure the channel will perform the intended Function.
The tested portion of the loop must trip within the Allowable Values specified in Table 3.3.2-1.Catawba Units 1 and 2 B 3.3.2-45 Revision NO@
ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
The setpoint shall be left set consistent with the assumptions of the setpoint methodology.
fl~~ ~~~~~~~~~
Iq~ 54O ~.,..C,, r 69~ T 2...Foy 5 CI P Ws 0A.: 1) WCAP-13900, "Extensio S of Slave Relay Surveilla.
SR 3.3.2.6 is the performance of a SLAVE RELAY TEST. The SLAVE Test Intervals," April 4; 2) RELAY TEST is the energizing of the slave relays. Contact operation is WCAP-13877 Re on 2-P-A, verified in one of two ways. Actuation equipment that may be operated in"Reliability As ssment of the design mitigation MODE is either allowed to function, or is placed in a Westingho e Type AR condition where the relay contact operation can be verified without Relays sed As SSPS Slave operation of the equipment.
Actuation equipment that may not be..
Revision 2, operated in the design mitigation MODE is prevented from operation by _"Reiablit Asessentof the SLAVE RELAY TEST circuit. For this latter case, contact operation is//" eiblty Asse s e to Potter & Brumfield MDR verified by a cont~inuity check of the circuit contaii~nq the slave relay _____.Series Relays," August 20 0. histet s omed every 92 d~yst' The Fre~iae ncy, is-adequat ,ae// Ion indus t/operating ex eriee conside ri 4'instrumnrc
,,.... Fr histor data. F" o/t--o-,o lv re.y orsaereasora raelynnthySA c ttataeo the auxiliary relays in the r e te ef so ESFAS circuit that are of Th westen r eib assedssmentts a re rr~felaysp i and apply only to the,9, the type Westinghouse Wetnh RadPte rmilMRtye relays. SSPS slave AR or Potter & Brumfield reayyo n axiiay.elysno adoedb MDR, the SLAVE RELAY qai fretne urveillance in rvals and will continue o be tested at TEST frequency is based a dyFeuny on operating experience, plant risk and is controlled SR 3.3.2.7 Surv nce Test Intervals," Apri V994, under the Surveillance o 1 Frequency Control SR 3.3.2.7 is the performance of a COT on the RWST level and in I P C i P A COT is performed on each required channel to ensure the entire channel will perform the intended Function.
Setpoints must be four within the Allowa Value ecified in Table 3.3.1-1. This test'performed ev 31 days. The Frequency is quate, base on operatingg gperience, co ng instru t reliab r I- i Catawba Units 1 and 2 B 3.3.2-46 Revision No.0 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.8 SR 3.3.2.8 is the performance of a TADOT. This test is a check of the Manual Actuation Functions, AFW pump start on trip of all MFW pumps, AFW low suction .pressure, Reactor Trip (P-4) Interlock, and Doghouse Water Level -High High Feedwater Isolation. cEach Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end device (i.e., pump starts, valve c cles etc.. ._"eq ncy s equ rase nindu oper g e ienc dis oisten th t cal r elin le The SR is modified by a Note thl des verification o setpoints dui gthe TADOT for manual initiatio unctions.
The manual initiation unctions have no associated.setpoint SR 3.3.2.9 I"J5_RT 2.SR 3.3.2.9 is the performance of a CHANNEL CALIBRATION.
aplimael ate"HANNEL CALIBRATION is a complete chec o the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter within the necessary range and accuracy.CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology.,Th F uency o 8 m onth bsed,,the assu ion of an 18 onth-lcali ation inte al in the termi~nat:V'n of he m afitu e o 'mn[.d rift in the__ _s epoi nt me This SR is modified by a Note stating that this test should include verification that the time constants are adjusted to the prescribed values where applicable.
The applicable time constants are shown in Table 3.3.2-1.SR 3.3.2.10 This SR ensures the individual channel ESF RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis.Response Time testing acceptance criteria are included in the UFSAR (Ref. 2). Individual component response times are not modeled in the Catawba Units 1 and 2 B 3.3.2-47 Revision No.0 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued) time could be affected is replacing the sensing assembly of a transmitter.
ESF RESPONSE TI tests are conducted on an 18 onth STAGGERED TES BASIS. Testing of the final uation devices, which make up the bul of the response time, is inclu d in the testing of each channel. The nal actuation device in one tr in is tested withi each channel. T refore, staggered testing re ts in response time Verificat!i/
of these devices every 18 rnenths. The 18 month Fre ency i3 cons tent with the typical refuelin zfycle and is based o~n u/" 0Pe ing experience, which show/hat r~andom failures of/i o~rumentation components cauing serious respns tin degradation, , utno channel, failure, are inf quent occurrence
ýThis SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 600 psig in the SGs.SR 3.3.2.11 SR 3.3.2.11 is the performance of a COT on the NSWS Suction Transfer-Low Pit Level.A COT is performed on each required channel to ensure the entire channel will perform the intended Function:
Setpoints must be found within the Allowable Values speciidi Table 3.3.2-1. -SR 3.3.2.12 .._,..... " --" SR 3.3.2.12 is the performance of an ACTUATION LOGIC TEST on the Doghouse Water Level-High High and NSWS Suction Transfer-Emergency Low Pit Level Functions.
An ACTUATION LOGIC TEST to satisfy the requirements of GL 96-01 is performed on each instrumentatio to ensur 1c combinations will iCbitiate the appropriate Function.
T s s1 3d ry eFqenis a u ate b n r .x Catawba Units 1 and 2 B 3.3.2-49 .Revision ESFAS Instrumentation B 3.3.2 BASES REFERENCES
- 1. UFSAR, Chapter 6.2. UFSAR, Chapter 7.3. UFSAR, Chapter 15.4. IEEE-279-1971.
- 5. 10 CFR 50.49.6. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 7. WCAP-1 0271-P-A, Supplement 1 and Supplement 2, Rev. 1, May 1986 and June 1990.8. WCAP-13632-P-A Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements" Sep., 1995.9. WCAP-14036-P-A Revision 1, "Elimination of Periodic Protection Channel Response Time Tests" Oct., 1998.10. CAP- 90, "Extension of ye Relay Surveillance T Inte als," April 1994.11. WCAP-1 3877 ision 2-P-A, "Reliability As ssment of Westingho Type AR Relays Used As PS Slave Relays," Augus 00.12. WCAP-13878-P-A Revision 2, liability Assessment o otter&rumfield MDR S ies Rela A 13. WCAP-14333-P-A, Revision 1, October 1998.(14. 5.13 1376-P-A,,01son 1 03 Catawba Units 1 and 2 B 3.3.2-50 Revision No-0 X PAM Instrumentation B 3.3.3 BASES SURVEILLANCE A Note has been added to the SR Table to clarify that SR 3.3.3.1 and REQUIREMENTS SR 3.3.3.3 apply to each PAM instrumentation Function in Table 3.3.3-1.SR 3.3.3.1 Performance of the CHANNEL CHEC evr3, da 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 something even more serious. 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 unit instruments located throughout the unit.Agreement criteria are determined by the unit 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. If the channels are within the criteria, it is an indication that the channels are OPERABLE.As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized.
hel Freq cyth oipaf 31 days is bas on operating exp ne that dmn ethat channel faie is rare. The C! NEL CHECK/pinslss formal, bifmore frequent, c becks of channels during ng rtoa s h i asasso ited with the tGOrequired' SR 3.3.3.2 Not Used Catawba Units 1 and 2 B 3.3.3-15 Revision No PAM Instrumentation B 3.3.3 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.3.3ý C LC LBRATI is perfogm nl r a r.imately -at eve _efu CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy.
This SR is modified by two Notes. Note 1 excludes neutron detectors.
The calibration method for neutron detectors is specified in the Bases of LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." Note 2 describes the calibration methods for.the Containment Area -High Range monitor. phperrequFicys bases4K, REFERENCES
- 1. UFSAR Section 1.8.2. Regulatory Guide 1.97, Rev. 2. , 3. NUREG-0737, Supplement 1, "TMI Action Items." 4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.3.3-16 Revision No.0 Remote Shutdown System B 3.3.4 BASES ACTIONS (continued)
B.1 and B.2 If the Required Action and associated Completion Time of Condition A is not met, the unit must be brought to. a MODE in which the LCO does not apply. To achieve this status, the unit 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 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 unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.3.4.1 REQUIREMENTS Performance of the CHANNEL CHECK ve 3/a nsures 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 the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation~continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff, based on a combination of the channel instrument.
uncertainties, including indication and readability.
If the channels are within the criteria, it is an indication that the channels are OPERABLE.
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 Surveillance, a CHANNEL CHECK is only required for those channels which are normally energized.
Tahe Fr ncy of ays is base1 on operati2 xperience we n dm sttrates th ichannel is rare. The CH ,K su lements I formal, but fore frequent hecks of cha rtels duringJ Siormal op eional use of~ dipaysi~n iasoted with theC requ--,ired.
Catawba Units 1 and 2 B 3.3.4-4 Revision No.O Remote Shutdown System B 3.3.4 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.4.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within' the necessary range and accuracy.The surveillance is modified by a Note that excepts the RTB position indication from a CHANNEL CALIBRATION.
The RTB position is indicated by a mechanical "flag" on the breaker.REFERENCES
- 1. 10 CFR 50, Appendix A, GDC 19.Catawba Units 1 and 2 B 3.3.4-5 Revision No.0 LOP DG Start Instrumentation B 3.3.5 BASES ACTIONS (continued)
B.1 Condition B applies when more than one loss of voltage or more than one degraded voltage channel on a single bus is inoperable.
Required Action B.1 requires restoring all but one channel to OPERABLE status. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time should allow ample time to repair most failures and takes into account the low probability of an event requiring an LOP start occurring during this interval.C.1 Condition C applies to each of the LOP DG start Functions when the Required Action and associated Completion Time for Condition A or B are not met.In these circumstances the Conditions specified in LCO 3.8.1, "AC Sources-Operating," or LCO 3.8.2, "AC Sources-Shutdown," for the DG made inoperable by failure of the LOP DG start instrumentation are required to be entered immediately.
The actions of those LCOs provide for adequate compensatory actions to assure unit safety.SURVEILLANCE REQUIREMENTS SR 3.3.5.1 SR 3.3.5.1 is the performance of a TADOT. i t islerfe a The test checks trip devices that provide actuation signals directly, bypassing the analog process control equipment.
For these tests, the relay Trip Setpoints are verified and adjusted as necessary.
![TheFreqji'ncyp based onrflthe knowniability of the rrpys and 996trols Testing consists-6T shedding and time le sensor relay testing only. Actuation of load timers is not required.Fr-or, LP/AsAlf, P e)7/o Catawba Units 1 and 2 B 3.3.5-5 Revision No. V)
Containment Air Release and Addition Isolation Instrumentation B 3.3.6 BASES 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 containment air release and addition, isolation Functions.
SR 3.3.6.1 SR 3.3.6.1 is the performance of an ACTUATION LOGIC TEST. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
In addition, the master relay coil is pulse tested for continuity.
This verifies that the logic modules are OPERAB E and there is an intact voltage signal path to the master relay coils._.Is es spror d every 92 ys on a ST E E,"fEST , ISAS e interval is 'tified i S R 3.3.6.3 ,441 W~ -7h Ie11Z SR 3.3.6.2 is the performance of a MAVTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage co d inuity check of the slave relay coil. Upon master relay.contact operation, a low voltage is injected to the slave relay coil. This voltage is insuffipient to pick upntoe stave relay, but large enough to demonstrate signal p th continuity.
T os tesate '-ev, Y29"d ays o na"* I I- IA1I[
i~nterval is ifed S R 3.3.6.-3 SIR 3.3.6.3 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation
- mode is either allowed to function or is placed in a condition where the relay contact operation can be verified without operation of the equipment.
Actuation*
equipment that may not be operated in the design mitigation mode is prevented
'from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuit check of the circuit containin the slave rela ase on tins meueliabi in stry eratin nth FoG ~FrA44A o2j Catawba Units 1 and 2 B 3.3.6-4 Revision No.1 I I .
Containment Air Release and Addition Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE REQUIREMENTS (continued)
- -u 1)WO pAP-1390 , 0. For slave r s or any auxiliary relays e ircuit that are of the type of SlaeReay Westingh seAR or Potter & Brumfie MDR" the SLAVE RELAY TES Test Interval April 1994;2) is perfor ed every 18 months. This est frequeno is based on the r 2 WCAP-1 7 Revision 2-P-A, reliabili assessments presented*
Referencehese"Relia ity Assessment of reliabs ty assessments are relay nd apply only to the s tiaose Type AR Relay Ws anghouse Type AR Relay .e inghouse AR an r rumfield MDR t e relays. SS S sla\Agst 3) WCAP- 878- rel s o xiliary relay not addressed b d not qua P-A Revision 2, "Relit ility extended surveillance' tervals and will continue to bet ted a 9, ro, r 5 Ff- r Awd" A'a -P, --,,ý /tA I/SR 3.3.6.4 WCAP-13900 ension of Slave Relay ellance Test Int. s," April 1994, For slave relays or any auxiliary relays in the circuit that are of the type Westinghouse AR or Potter &Brumfield MDR, the SLAVE RELAY TEST frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.SR 3.3.6.4 is the performance of a TADOT. This test is a check of the Manual Actuation Functions and is performed every 18 months. Each Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, etc.).The test also includes trip devices that provide actuation signals directly to the SSPS, bypassing the analog process control equipment.
The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.The Frequen, is ba on the known reliab -of the Functior r ancy avai le, and has been sln to be acceptaJ:ef operating e rience. /REFERENCES
- 1. 10 CFR 50.67.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
WCAP-13900, ension of Slave Relay Surveillance
...e...,.. In te rv a ls ,"'A 9 9 4 .,,. ./ ..4. WCAP- 877 Revision 2-P-A, "Reliability A ssment of W inghouse Type AR Relays Used as PS Slave Relays," d 2 Bugust 2 300. Ri 5. WCAP-13878-P-A Revision 2, eibltssessme f Potter&Brumfield MDR Series August 2000. i'WVCAP- 15376-P-A, viin 1", Mar;ch 2_003.i Id 2 B 3.3.6-5 Revision NO Catawba Units 1 an BDMS B 3.3.9 BASES ACTIONS (continued)
The Completion Times are based on the remaining OPERABLE BDMS train and the low probability of an event occurring during this time.B.1, B.2.1, B.2.2, B.3.1, and B.3.2 With both BDMS trains inoperable, the automatic capability for mitigation of dilution events is no longer available.
In this case, one BDMS train is required to be restored to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. As an alternative (Required Actions B.2.1 and B.2.2), operations involving positive reactivity additions must be suspended and valve NV-230 must be closed and secured within the following hour to isolate the unborated water sources. A third alternative (Required Actions B.3.1 and B.3.2) is to provide alternate methods of monitoring core reactivity conditions and controlling boron dilution incidents.
Alternative monitoring may be provided by the two Source Range Neutron Flux monitors.
These monitors must be verified to operate with alarm setpoints less than or equal to one-half decade (Square root of 10) above the steady-state count rate. In addition, the combined flowrate from both reactor makeup water pumps must be verified within the next hour to be within the limits specified in the COLR. Required Action B.2.1 is modified by a Note, which permits plant temperature changes provided the temperature change is accounted for in the calculated SDM and that keff remains <0.99. Introduction of temperature changes, including temperature increases when a positive MTC exists, must be evaluated to ensure they do not result in a loss of required SDM or adequate margin to criticality.
The Completion Times are based on the low probability of an event occurring during this time.SURVEILLANCE SR 3.3.9.1 REQUIREMENTS SR 3.3.9.1 is the performance of a CHANNEL CHECK on the BDMS, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels.
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, but each channel should be consistent with its cal The F~quenc E7of-i-2 houu.!i cosisncw e -CH EL)aECK FreUi 2ncy specified forR rostrume n LCO ...Catawba Units 1 and 2 B 3.3.9-3 Revision No.0' BDMS B 3.3.9 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.9.2 SR 3.3.9.2 is the performance of a COT for the BDMS, which is the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, display, and trip functions.
The COT also includes adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy. N3£'(This surveill ,,e must be pe frmed once days. The fre ency is based o erating exp nce, which ha own to be a~d ,01ate.SR 3.3.9.3 SR 3.3.9.3 is performed on the BDMS to verify the actuation signal causes the appropriate valves to move to their correct position and the Reactor Makeup Water Pumps stop to mitigate a boron dilution accident.The1 t rqenc s sd on the nee l,'o perform this /7 SSurvei nce under the c gdditions that ap ~furing a plant ou 'ge.Op tig experience, Js shown thes omponents usua fpass they S eillance when rormed at th 8 month Freque SR 3.3.9.4 SR 3.3.9.4 is the performance of a CHANNEL CHECK on the Source Range Neutron Flux monitors, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels.
It is IN56 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, but each channel should be consistent with its local conditions.
'A note is provided to clarify that the CHANNEL CHECK only needs to be performed on the Source Range Neutron Flux Monitors when used to satisfy Required Action A.3 or B.3.t a U netc y 1 a d 22 3 h3e9- 4 R e v isi o N o 0 Catawba Units 1 and 2 B 3.3.9-4 Revision Noo BDMS B 3.3.9 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.9.5 SR 3.3.9.5 verifies the combined flow rates from both Reactor Makeup Water Pumps are within the value specified in the COLR. This surveillance is only required when implementing Required Action A.3 or B.3. It ensures the assumptions in the analysis for the boron dilution event under these conditions are satisfied.
must be performed' conjunction with Re {A.3 or B.3 a j~~once per 31 days ,q is based on engine ig an te it kely event that a bo n dilution will occu I ing this.time.)-
SR 3.3.9.6 SR 3.3.9.6 is the performance of a COT for the Source Range Neutron Flux monitors, which is the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, display, and trip functions.
The COT also includes adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.
These monitors must beverified to operate with alarm setpoints less than or equal to 0.5 decade above the steady-state count rate. This SR is only required when the Source Range Neutron Flux Monitors are used to satisfy Required Action A.3 or B.3. This surveillance must be performed por to lacin the monitors in s ice for Reauired Action/ ,.3 or B.3 an qthereafter.47 da uncy stfie d f ___5 preC,,, 2. 10 CFR 50.36, Technicai Specifications, (c)(2)(ii).1 --A, Rev) 1, ,arc 3.Catawba Units 1 and 2 B 3.3.9-5 Revision No.
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES ACTIONS (continued)
RTP, the Power Range Neutron Flux -High Trip Setpoint must also be reduced to < 55% RTP. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to reset the trip setpoints is consistent with Required Action B.2. This is a sensitive operation that may inadvertently trip the Reactor Protection System.Operation is permitted to continue provided the RCS total flow is restored to > 99% of the value specified in the COLR within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is reasonable considering the increased margin to DNB at power levels below 50% and the fact that power increases associated with a transient are limited by the reduced trip setpoint.D.1 If the Required Actions are 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 />. 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 This surveillance demonstrates that the pressurizer pressure remains within the required limits. Alarms and other indications are available to aejteatrifthis limit is approached or exc~ced rquencyio recmanor in the control r lfor monitoring thd o ressure ant elated Iequip~ment status._ T 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval hýKebeen s~hown by oq ;rating i practi~ce to be su i_ ient to regularly ass9_6s for pote~ntialdea
!to verify oper ion is with In saft aPlysis _s,_p-i ihnsafety aO yss assump tions_SCR 3.4.1.2 This surveillance demonstrates that the average RICS temperature remains within the required limits. Alarms and other indications are e .available to alert operators if this limit is 2p~rahdo exedc.Th frequenc Of 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> s sufficient, considei-th oherr indic ýs \avail e_ to the o oorator in the control 'om for monitoring tl'e RICS. The ,12"hour interv rhas been shown by~perating practice fbe sufficient to Sregularly a ess for potenti ~ n~n o eration i iti safet a lss assumptions-
.Catawba Units 1 and 2 B 3.4.1-4 Revision Ndo-RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.1.3 This surveillance demonstrates that the RCS total flow rate remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded.
The frequen of 12 r! suficient, cMiseringjle other inaj aions availab to the t op or in the trol roo r monitori the RCS flow te and related'\quipment s us (e.g. R voltage a frequency,,.). ,he 1 or interval h been sho by operati practice to e sufficienjo regularly SR 3.4.1.4 Calibration of the installed RCS flow instrumentation permits verification that the actual RCS flow rate is greater than or equal to the minimum.required RCS flow rate. 5 4.(The Frequency 9)8~nths is consis ~erattxp§ ience.REFERENCES
- 1. UFSAR, Section 15.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.4.1-5 Revision Noir RCS P/T Limits B 3.4.3 BASES ACTIONS (continued)
Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the specified limits is requirecicoý C when RCS pressure nd-temperature conditions are .undegig planned changes..This Freq encyk_ coridere bleView elthe contl room in cation a yilable to pnitor R status./SAlso- ince tem, Frature ra teof specifie hr hourly ,].in 'ements, .minutes ~rmits ass.
j dvwations tfithin a rea onable time. __
.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. 10 CFR 50, Appendix G.2. ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.3. ASTM E 185-73, 1973 (Unit 1), E 185-82, 1982 (Unit 2).4. 10 CFR 50, Appendix H.5. Regulatory Guide 1.99, Revision 2, May 1988.6. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.7. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.4.3-6 Revision No-RCS Loops -MODES 1 and 2 B 3.4.4 BASES APPLICABILITY (continued)
LCO 3.4.5, "RCS Loops-MODE 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 3.4.17, "RCS Loops-Test Exceptions";
LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (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 3. 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 3 from full power conditions in an orderly manner and without challenging safety systems.SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verificationC IjIithat each RCS loop is in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is rovidinc val wleaintainin th ar to Freque of is suffiienle siderin er indicatioand alar avail e to ope or in the c rol room to m itor RCS op perfrmanc REFERENCES
- 1. UFSAR, Section 15.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.4.4-3 Revision RCS Loops -MODE 3 B 3.4.5 BASES ACTIONS (continued)
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 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.
RCP seal injection flow is not considered to be an operation involving a reduction in RCS boron concentration.
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 of coolant 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 criticality.
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.
Once the CRDMs have been de-energized by openingthe RTBs or de-energizing the MG sets, other methods to keep the CRDMs de-energized may be used. These methods are pulling fuses or opening sliding links in the rod control. cabinets.
This allows the flexibility for closing the RTBs or energizing the MG sets, while still preventing rod motion.SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verification<--I"h-ithat the required loops are in operation.
Verification includes flow rate, temperature, and pump status itorin which help ensure that forced flow is providing heat rerrnwv, The Fre cy of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />( tficient conId her indicand alar availabe to th;e<Derator in the cý ýtf61 room to mog*l65 RCS IPer~e1ormance.
Ihe SR 3.4.5.2 SR 3.4.5.2 requires verification of SG OPERABILITY.
SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is> 12% for required RCS loops. If the SG secondary side narrow range water level is < 12%, the tubes may become uncovered and the associated loop may not be capable of providina the heat sink for Cedb heUt 1 ha2reque i consB4 adeqvision-Noother in lr in ions lp 5ol rooK alert the,- /.Slosso' level C aiawba Units 1 and 2 B 3.4.5-5 Revision No"'
RCS Loops -MODE 3 B 3.4.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.5.3 Verification that the required RCPs are OPERABLE ensures that safety analyses limits are met. The requirement also 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 availability to the required RCPs.REFERENCES
- 1. 10 CFR 50.36, TechnicalSpecifications, (c)(2)(ii).
Or)$1 Catawba Units 1 and 2 B 3.4.5-6 Revision No/z'--7J RCS Loops -MODE 4 B 3.4.6 BASES ACTIONS (continued)
C.1 and C.2 If no loop is OPERABLE or in operation, except during conditions permitted by Note 1 in the LCO section, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1 and maintain keff < 0.99 must be suspended and action to restore one RCS or RHR loop to OPERABLE status and operation must be initiated.
RCP seal injection flow is not considered to be an operation involving a reduction in RCS boron-concentration.
The required margin to criticality must not be reduced in this type of operation.
Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 and maintain kef < 0.99 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 and ke, requirements maintains acceptable margin to criticality.
The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.
SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verificationr/12ours)that one RCS or RHR loop is in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ens...ure that forced flow is prnvidinnh.a rpmval J'he Freque~nnp of 1t2 hour,%s sufficient c g,,ring other c no ]alrm to t o~perator in jtbe 'ntoý ý r Snnt, nitor Z__nýLRH;4oo p prfor .SR 3.4.6.2 SR 3.4.6.2 requires verification of SG OPERABILITY.
SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is> 12%. If the SG secondary side narrow range water level is < 12%, the tubes may become uncovered and the associated loop may not be ca able of providing the heat sink necessary for removal of decay hg.t-T 2h ureuncy is consid .ed adequa e Fin ew of other Catwicati°U con1ro ad 2e art Be :4ra.t r t os Catawba Units 1 and 2 B 3.4.6-4 Revision Nol" RCS Loops -MODE 4 B 3.4.6 BASES SURVEILLANCE REQUIREMENTS (continued)
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 ver in--roper breaker alignment and ower a.ailable to the re uired pump..The irýquency of.3ys is co idered reonabi view of ramns acontroe a has bee -h bown REFERENCES
- 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.4.6-5 Revision Noo RCS Loops -MODE 5. Loops Filled B 34.7 BASES ACTIONS A.1 and A.2 If one RHR loop is inoperable and the required SGs have secondary side narrow range water levels < 12%, redundancy for heat removal is lost Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore the required SG secondary side water levels. Either Required Action A.1 or Required Action A.2 will restore redundant heat removal paths. The immediate Completion Time reflects.the importance of maintaining the availability of two paths for heat removal.B.1 and B.2 If no RHR loop is in operation, except during conditions permitted by Note 1, or if no loop is OPERABLE,.all operations involving introduction of coolant into the ROS with boron concentration less than required to meet SDM of LCO 3.1.1 must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated.
RCP seal injection flow is.not considered to be an operation involving a reduction in RCS boron concentration.
Suspending the introduction of coolant into the RGS of coolant with boron concentration less than required to meet the minimum SOM 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 criticality.
The immediate Completion Times reflect the importance of maintaining operation for heat removal.SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification v Iorat the required loop is in operation.
Verification includes I1ow-r-af-e, temperature, or pump status monitorin , which help ensure that forced flow s vn at r oval.'e Fr- e'nc--y 12 o o ufficient dering rdi ictin -nd ala availa orator in -mcotrolmrortomnt4Rlop,SR 3.4.7.2 Verifying that at least two SGs are OPERABLE by ensuring their secondary side narrow ranqe water levels are >_ 12% en 4 i!res an tqfrr'!-decay heat removal method in the event that the second RHR loop is not OPERAB th RHR loos are OPERABLE, this Surveillance is not needed The o -Freque is co edred equate iaew of o r din lleons ieinth antro, in to al the ooe or toth ssof Catawba Units 1 and 2 B 3.4.7-4 Revision No(2 RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.7.3 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 RHR pump.If secondary side narrow range water level is > 12% in at least two SGs, this Surveillance is not needed. Th F uency of 7a is consi reasonable in- v of o er admini tive as been.shown to Jaeacceptable by op ti ng e REFERENCES
- 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.4.7-5/A Revision Noo RCS Loops -MODE 5, Loops Not Filled B 3.4.8 BASES ACTIONS (continued) immediately to restore an RHR loop to OPERABLE status and operation.
RCP seal injection flow is not considered to be an operation involving a reduction in RCS boron concentration.
The required margin to criticality must not be reduced in this type of operation.
Suspending the introduction of coolant into the RCS of coolant 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 criticality.
The immediate Completion Time reflects the importance of maintaining operation for heat removal.The action to restore must continue until one loop is restored to OPERABLE status and operation.
SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification 12 urs that one loop is in operation.
Verification includes flow rate, temperture, or pump status monitoring, which hel ensure that forced flow is providing heat removal. be reqenc f 112:ý hoF r ( suff ; .t sidTe-r, n-g`-omURrndicatio'n .nd!alarms vail control rX fnto monit H'R l-- oop.SR 3.4.8.2 Verification that the required number of pumps are 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 ro per breaker alignment and power availabla-, _t required pumps., he Frouencyo days is clsidered r asonajie in view of oter administratiWe contr(il abl nd has be6n be acc able by operatio experihnc REFERENCES
- 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.4.8-3 Revis on No Pressurizer B 3.4.9 BASES ACTIONS (continued)
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 steamb l. The Surveillance is performed by observing the indicated requency of th rs corresponds preparam r heatesitt The 12 h i terval has been sh oerating.This prS me tov , sufclient to r jlarly assess level for alleviation and v that avia o arl, detection of abno rmal level indicatol SR 3.4.9.2 The SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power and the associated pressurizer heaters are verified to be at their design rating. This SR ma ma. be verified , by energizini the heaters and measuring cyurrent.
e re 9 si cn'ed adequate to [tet heatr radatiFnd 2 1CR0n shown .perating expeia to be This Surveillance demonstrates tha t the heaters can be transferred from the no to the emergency power supply 6~frequ 'c-y of 18 is baseado, tyi5c~al ,i coZnsi~et w wit iir ctoso gc oi milar ve ications of emerg cy power REFERENCES 1 .UFSAR, Section 15.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 3. NUREG-0737, November 1980.Catawba Units 1 and 2 B 3.4.9-4 Revision No.
Pressurizer PORVs B 34.11 BASES ACTIONS (continued)
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. In MODES 4 and 5, maintaining PORV OPERABILITY may be required.
See LCO 3.4.12.SURVEILLANCE REQUIREMENTS SR 3.4.11.1 / 5 ... ..Block valve ccling verifies that the valve(s) can be closed if n ede sWb fo cyy of 92 Sis t..t, .Coe If the b-iot-alve is closed to isofae a PORV that is c bef b~ing manually cycled,,the OPERABILITY of the block valve is of importance, because*opening the block valve is necessary to permit the PORV to be used for.manual control of reactor pressure.
If the 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 wei. athe allowable limits (25%) to extend the block valvef requenc I f Furthermore, these test requirements would be complete"y 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 met with the block valve closed, in accordance with the Required Action of this LCO.SR 3.4.11.2 SR 3.4.11.2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORVcan be manually actuaed for mitiaation of an SGTR.oTheoFnof 1 n />ed pical refue cycle a stry ed pr e The SR is modified by a Note which states that the SR is required to be performed in MODE 3 or 4 when the temperature of the RCS cold legs is> 200OF consistent with Generic Letter 90-06 (Ref. 5).Catawba Units 1 and 2 B 3.4.1 1-6 c7lýRevision No.
Pressurizer PORVs B 3.4.11 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.11.3 The Surveillance demonstrates that the emergency nitrogen supply can be provided and is performed by transferring power from normal air supj to emergency nitrogen supply and cycling the valves. jThe Frequoiyot This SR is modified by a Note which states the SR is not applicable to PORV NC-36B. This PORV does not have a nitrogen supply.REFERENCES
- 1. Regulatory Guide 1.32, February 1977.2. UFSAR, Section 15.4.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 4. ASME Code for Operations and Maintenance of Nuclear Power Plants.5. 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," Pursuant to 10 CFR 50.54(f) (Generic Letter 90-06).Catawba Units 1 and 2 B 3.4.11-7 Revision No-D LTOP System B 3.4.12 BASES SURVEILLANCE SR 3.4-12.1 and SR 3.4.12.2 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, a maximum of two pumps (charging and/or safety injection) are verified capable of injecting into the RCS and the accumulator discharge isolation valves are verified closed and power removed.The 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 two valves in the discharge flow path being closed.Th requenc 12- h s sufficient, cd-ering other tions and arms av le to t erat in the'control roo verifv the reif'ed statu the equi ment.SR 3.4.12.3 Each required RHR suction relief valve shall be demonstrated OPERABLE by verifying its RHR suction isolation valves are open and by testing it in accordance with the Inservice Testing Program. This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO.The RHR ction isolation valves are verified to be openedrJ-g v ~~.The Fr_ uenc is co dere e in view of ther mi stra e controls s as valve atus indication available to ,o l, operato n the contr o hatveif he RHR su ion isolatio alve The ASME Code (Ref. 9), test per Inservice Testing Program verifies OPERABILITY by proving relief valve mechanical motion and by measuring and, if required, adjusting the lift setpoint.SR 3.4.12.4 The PORV block valve must be verified open e.-. rqto provide the flow path for each required PORV to perfom i s unc i when actuated.
The valve must 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 Catawba Units 1 and 2 B 3.4.12-11 Revision No4 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued) power to the valve operator is not required removed, and the manual operator is not required locked in the inactive position.
Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation.
Th 2 our Frequency is co ered adequate 'i~ae fote/SR 3.4.12.5, ,,!df L- ,$ i:. 3. '. IZ. 2 Performance of a COT is required ithin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after decreasing RCS temperature to < 210°F an he on each required PORV to a,,.n--Y setpoint is within the allowed maximum limits. PORV actuation could tdepressurize the RCS and is not required.
r P t tThe 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency considers the unlikelihood of a o pa ure overpressure event during this time.Fay 5F7C P A Note has been added indicating that this SR isrequired to be met 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 < 210 0 F. The COT cannot be performed until in the LTOP MODES when the PORV lift setpoint can be reduced to the LTOP setting. The test 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.SR 3.4.12.6 Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required ev(tto adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known input.S.R 3.4.12.7 Each required RHR suction relief valve shall be demonstrated OPERABLE by verifying its RHR suction isolation valves are open and by testing it in accordance with the Inservice Testing Program. (Refer to SR 3.4.12.3 for the RHR suction isolation valves Surveillance and for a description of the Inservice Testing Program.)
This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO.Catawba Units 1 and 2 B 3.4.12-12 Revision No-ZI)
LTOP System B 3-4.12 BASES SURVEILLANCE REQUIREMENTS (continued)
RHR suction isolation valves are verified open, with power to the valve operator removed and locked in the removed position, to ensure that accidental closure will not occur. The "locked open in the removed position" power supply must be locally verified in its open position with the power supply to the valve locked in its inactive position.J he 31 day Fr ýency is based o ineering ju. tnt, is consist} with the p ýeedural control v~lv oiration, and e es (corre ,alve position '.j- //tlgdT -I- )REFERENCES
- 1. 10 CFR 50, Appendix G.2. Generic Letter 88-11.3. UFSAR, Section 5.2 4. 10 CFR 50, Section 50.46.5. 10 CFR 50,,Appendix K.6. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 7. Generic Letter 90-06.8. ASME, Boiler and Pressure Vessel Code,Section III.9- ASME Code for Operation and Maintenance of Nuclear Power Plants.Catawba Units 1 and 2 B 3.4.12-13 Revision No/"--Z-1)
V11--j RCS Operational LEAKAGE B 3.4.13 BASES'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 would at first appear as unidentified LEAKAGE and can only be positively identified by inspection.
It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and Identified LEAKAGE are determined by performance of an RCS water Inventory balance. For this SR, the volumetric calculation of unidentified LEAKAGE and identified LEAKAGE Is based on a density at room temperature of 77 degrees F.The Surveillance is modified by two Notes. The RCS water inventory balance must be performed with the reactor at steady state operating conditions and near operating pressure.
Therefore, Note 1 indicates that this SR is not required to be completed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of steady state operation near operating pressure have been established.
Steady state operation is required to perform a proper inventory balance;calculations during maneuvering are not useful and Note 1 requires the Surveillance to be met when steady state is established.
For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RGS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letcdown, and RCP seal iniection and return flows.Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day or lower cannot be measured accurately by an RCS water inventory balance.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. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation.", The hour Frequen s a reasonable in alto trend LEA and! ognizes the im rtance of early le age detecti n evnti of ccidents an duction of pote econseauence A Note under the requency column states that this SR is nly required to be performed during steady state operation.
Catawba Units 1 and 2 B 3.4.13-5 Revision N60 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less than 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 3.4.18, "Steam Generator (SG) Tube Integrity," should be evaluated.
The 150 gallons per day limit is based on measurements taken at room temperature.
The primary to secondary leak rate assumed in the safety analyses is taken also at room temperature.
The Surveillance is modified by a Note which states that this SR is not required to be completed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of steady state operation near operating pressure have been established.
During normal operation the primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling.e 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> equency is a reasona e-lnterval to trend pr ary to second EAKAGE and reco gades the importanc early leak e det ion in th reventior eaccidents and redu ion of ote nsequences.
Note und6r the Frequency column states that this SR is only require to e.perfo duxa,,steady state operation.
REFERENCES
- 1. 10 CFR 50, Appendix A, GDC 30.2. Regulatory Guide 1 .45, May 1973.3. UFSAR, Section 15.4. 1-&-CFR 50.36, Technical Specifications, (c)(2)(ii).
- 5. EPRI TR-104788-R2, "PWR Primary-to-Secondary Leak Guidelines," Revision 2.6. NEI 97-06, "Steam Generator Program Guidelines." 7. UFSAR, Section 18, Table 18-1.8. Catawba License Renewal Commitments, CNS-1274.00-00-0016, Section 4.27.9. 10 CFR 50.67.Catawba Units 1 and 2 B 3.4.13-6 Revision No RCS PIV Leakage B 3.4.14 BASES ACTIONS (continued)
B.1 and B.2 If leakage cannot be reduced, or the other Required Actions accomplished, 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 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 and also reduces 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.C.1 The RHR interlock prevents the RHR suction isolation valves inadvertent opening at RCS pressures in excess of the RHR systems design pressure.
If the RHR interlock is inoperable, operation may continue as long as the affected RHR suction penetration is closed by at least one closed manual or deactivated automatic valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This Action accomplishes the purpose of the interlock function.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 is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition.
For the two PIVs in series, the leakage requirement applies to each valve individually 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 valve in series meets the leakage requirement.
In this situation, the protection provided by redundant valves would be lost.Tes i'sto be performed eve q 18 months, a typical refeg cycle, if the 1nt does not go into M 5 for at least 7 days. e 18 month Frequency is cons tnt with 10 CFR 50.55a( ef. 9) as contained in the InserviS ting Program, is within jruency allowed by the Amen Society of Mechanical En eers (ASME) Code,. 8), and is based on the need to perform s surveillances und e conditions that Catawba Units 1 and 2 B 3.4.14-4 Revision No0 RCS PIV Leakage B 3.4.14 BASES SURVEILLANCE REQUIREMENTS (continued)an o ge and the pot lral for an un 4nned cere performed tfhhe reactor p"ower..--
I/VSL r In addition, testing must be performed once after the valve has been opened by flow or exercised to ensure tight reseating.
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.
Within 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.
The Note that allows this provision is complementary to the Frequency of prior to entry into MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months. In addition, this Surveillance is not required to be performed on the RHR System when the RHR System is aligned to the RCS in the shutdown cooling mode of operation.
PIVs contained in the RHR shutdown cooling flow path must be leakage rate tested after RHR is secured and stable unit conditions and the necessary differential pressures are established.
SR 3.4.14.2 Verifying that the RHR interlock is OPERABLE ensures that RCS pressure will not pressurize the RHR system beyond its design pressure of 600 psig. The interlock setpoint that prevents the valves from being opened is set so the actual RCS pressure must be < 425 psig to open the valves. This setpoint ensures the RHR design pressure will not be exceeded and the RHR relief valves will not lift. The month Fre ncy nis bas onBthe to pe3 4 m 5 the Sueillance der conditio hat I he 8i, ronth Fre ency is durin con rut min Thein 1 Catawba Units 1 and 2 B 3.4.14-5 Revision No. V RCS Leakage Detection Instrumentation B 3.4.15 BASES ACTIONS (continued)
G.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.
The required monitors during MODE 1 for LCO 3.0.3 entry are defined as the simultaneous inoperability of one CFAE level monitor, the containment atmosphere particulate radioactivity monitor, and the CVUCDT level monitor. The required monitors during MODES 2, 3, and 4 for LCO 3.0.3 entry are defined as the simultaneous inoperability of one CFAE level monitor and the CVUCDT level monitor.This condition does not apply to the incore instrument sump level alarm.SURVEILLANCE SR 3.4.15.1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the containment atmosphere particulate radioactivity monitor. The check gIves reasonable confidence that the channel is operating Fre-ncy 12 hou s base n instr nt re iai an blej fdet 4fng off 10a o a SR 3.4.15.2 SR 3.4.15.2 requires the performance of a COT on the containment atmosphere particulate radioactivity monitor. The test ensures that a signal from the monitor can generate the appropriate alarm associated with the detection of a minimum 1 gpm RCS leak. The desired alarm is derived from a digital database.
Database manipulation concurrent with a signal supplied from the detector verifies the OPERABILITY of thedayd ah o'nsiderfT9strumentIr~ability, i.nd ting expei.enc has,4hown it is_ detectig ,, SR 3.4.15.3, SR 3.4.15.4, SR 3.4.15.5, and SR 3.4.15.6 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 inCstruments loc.ated-inside containment.
T1 anreency o2 14.smonthsoa-iiclrru' g cycle an nnel re .bifity. ýAga i has proven this F queency is Catawba Units 1 and 2 .....' F ,ievision N
RCS Specific Activity B 3.4.16 BASES ACTIONS (continued) transient specific activity excursions while the plant remains at, or proceeds to power operation.
.. B.1 With the gross specific activity in excess of the allowed limit, the unit must be placed in a MODE in which the requirement does not apply.The change within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to MODE 3 and RCS average temperature
< 500°F lowers the saturation pressure of the reactor coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment in an SGTR event. 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, to reach MODE 3 below 500°F from full power conditions in an orderly manner and without challenging plant systems.C.1 If a Required Action and the associated Completion Time of Condition A is not met or if the DOSE EQUIVALENT 1-131 is in the unacceptable region of Figure 3.4.16-1, the reactor must be brought to MODE 3 with RCS average temperature
< 500°F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. 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 3 below 500 0 F from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolanp aflea t on: ee CA gross radioactivity analysis shall consist _o e uantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half-lives less than 10 minutes and all radioiodines.
The total specific activity shall be the sum of the beta-gamma activity in the sample within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the sample is taken and extrapolated back to when the sample was taken. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level. The latest available data may be used for pure beta-emitting radionuclides.
This Surveillance provides an indication of any increase in gross specific activity.Catawba Units 1 and 2 B 3.4.16-4 Revision Nc(
RCS Specific Activity B 3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued)
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 a plicable in MODES 1 and 2, and in MODE 3 with Tavg at least 500 0 F. The F req uep onsid ,::!e unlikelihooof a"f6s-~-lt6"Tf'"aire 6-du-nna t~re I--r. -ý /I-'-SR 3.4.16.2 This Surveillance is performed in MODE 1 only to ensure iodine remains within limit during normal operation ancd following fast power changes when fuel failure is more apt to occur. 'h. 14 requency I, euao tnd ch rres i e iodine ., vi ty n' irerinn n."rIvityJ j ,N eve days'. he Frequency, between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> a er a poe-r 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. If the power excursion is one continuous process spanning over several hours, there is no need to sample every hour, only 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the last major power change of> 15% RTP, since this sample will encompass the maximum potential for additional iodine release to have occurred.SR 3.4.16.3 A radiochemical analysis for E determination is required Cgwith the plant operating in MODE 1 equilibrium conditions.
The E determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit. The analysis for E is a measurement of the average energies per disintegration for istpswith half lives longer than 10 minutes, excluding Fr ync 184 d 'ec n*, -- s Er,-o er This SR has been modified by a Note that indicates sampling is required-to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed 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 />. 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.Catawba Units 1 and 2 B 3.4.16-5 Revision N 6-ý RCS Loops -Test Exceptions B 3.4.17 BASES SURVEILLANCE REQUIREMENTS SR 3.4.17.1 Verification that the power level is < the P-7 interlock setpoint (10%) will ensure that the fuel design criteria are not volated durinn the performance of the PHYSICS TESTS. IThe Frequenc once per ho s ad equate to e qre that the po level does noexceed the li.Plant oper is are conducte owly during th erformancef PHY TESTS and mon' ring the power el once per our is s icient to ensure that t e power level d s not exceedthe limit. 'SR 3.4.17.2 The power range and intermediate range neutron detectors and P-1 0 and P-1 3 inputs to the P-7 interlock setpoint must be verified to be OPERABLE and adjusted to the proper value. A COT is performed prior to initiation of the PHYSICS TESTS. This will ensure that the RTS is properly aligned to provide the required degree of core protection during the performance of the PHYSICS TESTS.REFERENCES
- 1. 10 CFR 50, Appendix B, Section XI.2. 10 CFR 50, Appendix A, GDC 1, 1988.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.4.17-3 Revision NoO Accumulators B 3.5.1 BASES SURVEILLANCE REQUIREMENTS SR 3.5.1.1 Each accumulator valve should be verified to be fully openl TThis verification ensures that the accumulators are available or njection 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, d valve could result in not meeting accident analyses assumptions qrc n s CO eredin view o er admis ra on I at ensur mi,(ositioned ation valve iu nlikely.SR 3.5.1.2 and SR 3.5.1.3 Qý ýýorated water volui verified for each accumulator.
This installed control room indication.
adequate injec n during a LO accumulato , a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Fr ency i change efore limits reached.SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulator since the static design of the accumulators limits the ways in which the concentration can be changed.h_31 d~a 1eec ii; a "e6quat ident i, y ~es tha~tould u anisms such a atification o akaeC.qmplin the- -_affected accumulator within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> at er a 75 gallon increase will identify whether inleakage has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST), because the water contained in the RWST is within the accumulator boron concentration requirements.
This is consistent with the recommendation of NUREG-1366 (Ref. 7).Catawba Units 1 and 2 B 3.5.1-7 Revision No.
Accumulators B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.5 Verification e3a'that power is removed from each accumulator isolation valve operators for N154A, N165B, N176A, and N188B when the RCS pressure is > 1000 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, only two accumulators would be available for injection given a single failure coincident with a LOCA.'e -owe r is emoved a~l circui kers doc-ked lnder/control, t l~"31 da will, pro e adeg{uate a uranc hat powe " remove .-This SR allows power to be supplied to the motor operated isolation valves when RCS pressure is < 1000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns.
Even with power supplied to the'valves, inadvertent closure is prevented by the RCS pressure interlock associated with the valves.Should closure of a valve occur in spite of the interlock, the SI signal provided to the valves would open a closed valve in the event of a LOCA.REFERENCES
- 1. IEEE Standard 279-1971 2. UFSAR, Chapter 6.3. 10 CFR 50.46.4. DPC-NE-3004.
- 5. 10 CFR 50.36, Technical Specification, (c)(2)(ii).
- 6. WCAP-15049-A, Rev. 1, April 1999.7. NUREG-1366, February 1990.Catawba Units 1 and 2 B 3.5.1-8 Revision No.J ECCS -Operating B 3.5.2 BASES ACTIONS (continued)
An event accompanied by a loss of offsite power and the failure of an EDG can disable one ECCS train until power is restored.
A reliability analysis (Ref. 6) has shown that the impact of having one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.Reference 7 describes situations in which one component, such as an RHR crossover valve, can disable both ECCS trains. 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.B.1 and B.2 If the inoperable trains 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.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.
Misalignment of these valves could render both ECCS trains inoperable.
Securing these valves using the power disconnect switches in the correct position ensures that they cannot change position as a result of an active failure or be inadvertently misaligned.
These valves are of the type, described in Reference 7, that can disable the function of both E t anal yses. 2 r Freoncy is,66onsidprdldrgeaso a~le in. of ot-=" j~l"~'~Is r f e t Will ~nsure.,A'mispos'ined va~be i nl SR 3.5.2.2 /ii.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, Catawba Units 1 and 2 B 3.5.2-7 Revision No.
ECCS -Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued) 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 th se valves ca ble of being mispositioned are in the correct osition. he f'1 d_ Frequu ncy i ppro iate becaus he valves are dperated under a inistra'e control. .F)queency ha 'reen sh.own, be accept through n pe'ating i SR 3.5.2.3 With the exception of the operating centrifugal charging pump, the ECCS pumps are normally in a standby, nonoperating mode. As such, flow path piping has the potential to develop voids and pockets of entrained gases.ECCS piping is verified to be water filled by venting to remove gas from accessible locations susceptible to gas accumulation.
Alternative means may be used to verify water filled conditions (e.g., ultrasonic testing or high point sight glass observation).
Maintaining the piping from the ECCS pumps to the RCS full of water ensures that the system will perform properly, injecting its full capacity into the RCS upon demand.This will also prevent water hammer, pump cavitation, and pumping of noncondensible gas (e.g., air, nitrogen, or hydrogen) into vessel following an SI signal or during shutdown cooling .The 31 d " reFqu cy taKA into considieation the~ejradual n of gas _Sacc amulatio in the a ndthe ..roc idural *ontro, s te m o ;rat io°n. SR 3.5.2.4 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code. This type of testing may be accomplished by measuring the pump developed head at only one 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, the ASME Code.The ASME Code provides the activities and Frequencies necessary to satisfy the requirements.
Catawba Units 1 and 2 B 3.5.2-8 Revision ECCS -Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.5 and SR 3.5.2.6 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI and Containment Sump Recirculation 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 oth.red'in the reqguuired a~mini2rative controls./The
.18 nth Frequency,/
baern teed to prfor te Surveillances the conditio s (that apply d/zring a plant outageý tPnd the potential o yunplanned plan/trans~ient~sif the Surveillances/
ere performed wi the reactor at The 18 Znth Frequency i also acceptable b ~ed on considera/
on of the sign reliability (an onfirming operati experience) of e e ipment..The actua tion logic is part of ESF A.ctt tion./1ystemn testing, and equipmetef nc smnitore s ppart of the /
ror.SR 3.5.2.7 The position of throttle valves in the flow path on an Sl signal is necessary for proper ECCS performance.;
These valves have mechanical locks to ensure proper positioning for restricted flow to a ruptured cold leg, ensuring that the other cold legs receive at lea~st the required minimum flow. oncy-I s, e-r'-a-, dns.SR 3 .2.5 a oa R 3. ..6.]1 ; -SR 3.5.2.8 Periodic inspections of the containment sump suction inlet n ure that it is .unrestricted anda sta s ia ý,oJ2er operating oditio .TLhe I 4"onth/" (Freque is base on theud to perf~m thisS veilian under e condt ns that ply dur g a plantg age an n the ed to e aeSS to th. ocation. has 'bee n d_ o5-e teo degaatio, is conir e'y operati; experi Upon completion of the ECCS sump strainer assem y modifications during outage 2EOC15 for Unit 2 and 1EOC17 for Unit 1, the following SR Bases will apply: Periodic inspections of the ECCS containment sump strainer assembly (consisting of modular tophats, grating, plenums, and waterboxes) ensure it is unrestricted and remains in proper operating condition.
Catawba Units 1 and 2 B 3.5.2-9 Revision No NO CHANGES THIS PAGE.FOR INFORMATION ONLY ECCS -Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
Inspections will consist of a visual examination of the exterior surfaces of the strainer assembly for any evidence of debris, structural distress or abnormal corrosion.
The intent of this surveillance is to ensure the absence of any condition which could adversely affect strainer functionality.
Surveillance performance does not require removal of any tophat modules or grating, but the strainer exteriors shall be visually inspected.
This surveillance is not a commitment to inspect 100% of the surface area of all tophats, but a sufficiently detailed inspection of exterior strainer surfaces is required to establish a high confidence that no adverse conditions are present. The scope of inspection necessary to provide high confidence includes 100% of the strainer areas that can be accessed and inspected using normal means and tools (i.e., flashlight, extendable mirror, hand held digital camera) without disassembly, and that difficult to access areas will be inspected to the extent possible using these same means.Any damage detected in the strainer assembly inspection will result in an expansion of the scope of the inspection to include other areas of potential damage. Inspection scope should be expanded, as needed, for degradation of strainer components identified during this inspection that were not considered readily accessible during the inspector's initial evaluation.
REFERENCES
- 1. 10 CFR 50, Appendix A, GDC 35.2. 10 CFR 50.46.3. UFSAR, Section 6.2.1.4. UFSAR, Chapter 15.5. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 6. NRC Memorandum to V. Stello, Jr., from R.L. Baer,"Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.7. IE Information Notice No. 87-01.Catawba Units 1 and 2 B 3.5.2-10 Revision No I1 RWST B 3.5.4 BASES ACTIONS (continued)
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 borated water temperature should be verifie .ve the to be within the limits laseumeder to ensretht anal s u a l supl aFreq fncrs sinjient to ispontify a teorature E ge thainoumd to -accept ethr.oug-h ,C jlw( op r a0t&ing perienc S FCt Fo,. SI-cP C*) SR 3.5.4.2 The RWST water volume should be verified 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 Containment Spray System pump operation on recirculation.
jnc Rv'l vo ef"'innonally sw le and i protec y a avalarmb ae ayhFrieqsueres tha t ropria sn sp n to an ac eptarluge erating stJ S R 3.5.4.3 The boron concentration of the RWST should be verifieddtoA&
be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA and that the boron content assumed for the injection water in the MSLB analysis is available.
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 svstems-andmcomponen
-Mm Catawba Units 1 and 2 B 3.5.4-5 r7\Revision No.(ý)
Seal Injection Flow B 3.5.5 BASES ACTIONS (continued) operator has 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from the time the flow is known to be above the limit to correctly position the manual valves and thus be in compliance with the accident analysis.
The Completion Time minimizes the potential exposure of the plant to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow within limits.This time is conservative with respect to the Completion Times of other ECCS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel.
B.1 and B.2 When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be initiated.
The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators.
Continuing the plant shutdown begun in Required Action B.1, an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.
SURVEILLANCE REQUIREMENTS SR 3.5.5.1 Verification
ýý3ýthat the manual seal injection throttle valves are adjusted to give a flow within the limit ensures that proper manual seal injection throttle valve position, and hence, proper seal injection flow, ,is maintained.
..he Frec of ýday's is base .on enginee -rg " udgen d is. c sistn wt;hr ECC.S .lve ce SFrequ ,tcies. Tfe Fre rov 0 be accept 1e through...
tig ..As noted, the Surveillance is required to be performed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the RCS pressure has stabilized within a +/- 20 psig range of normal operating pressure.
The RCS pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that the manual valves are set correctly.
The exception is limited to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to ensure that the Surveillance is timely.Catawba Units 1 and 2 B 3.5.5-3 Revision No. 6 Containment Air Locks B 3.6.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.2.2 Door seals must be tested so verify the integrity of the inflatable door seal. The leakage rate must be less than 15 standard cubic centimeters per minute (sccm) per door seal when the seal is inflated to approximately 85 psig. This ensures that the seals will remain inflated for at least 7 days should the instrument air supply to the seals be lost. ofe~stiA ed sra 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 o ening of the inner and outer doors will not inadvertently occur Fue to thei'"purely rnec lnical of this interlo~ef{
and given t-hat the inter ck/mechaniz is not normally challengeý when the containment a' lock Idoor i sed for entry and exit (pr edures require strict adheence todoor opening), this test is/ nly required to be proied every 18 -! rnths. The 18 month Frequ ency is based on the nee , perform this!/urveillance und'er the con li ions that apply during a ;ant outage, and Sthe potential for Iloss of c tainment OPERABILI-IYA
/t the surveillance were performed wIt ti reactor at power.-The Frequency for the interlock is justif d based on gen( eri oprng experience. Frequency is bas on engineering judgme and is considered adequate givenCthatathwbotchalainged d ring the use of the interlock-.
Catawba Units I and 2 B 3.6.2-7 Revision No.0 Containment Isolation Valves B 3.6.3 BASES ACTIONS (continued)
For the isolation devices inside containment, the time period specified as"prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of the inaccessibility of the isolation devices and other administrative controls that will ensure that isolation device misalignment is an unlikely possibility.
For the valve with resilient seal that is isolated in accordance with Required Action E.1, SR 3.6.3.6 must be performed at least once every 92 days. This assures that degradation of the resilient seal is detected and confirms that the leakage rate of the containment purge valve does not increase during the time the penetration is isolated.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 Each containment purge supply and exhaust isolation valve for the lower compartment and the upper compartment, instrument room, and the H dropen Purge System is required to be verified sealed closed at y This Surveillance is designed to ensure that a gross...breacf-containment is not caused by an inadvertent or spurious opening of a containment purge valve. Detailed analysis of these valves to conclusively demonstrate their ability to close during a LOCA in time to limit offsite doses has not been performed.
Therefore, these valves are required to be in the sealed closed position during MODES 1, 2, 3, and 4.A valve that is sealed closed must have motive power to the valve operator removed. This can be accomplished by de-energizing the source of electric power or by removing the air supply to the valve operator.
In this application, the term "sealed" has no connotation of leak tightness.
Catawba Units 1 and 2 B 3.6.3-10 Revision No...
Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)
'-3 he Fre "sncys a resu an NRC iitn e, Generic I -24 related to c ,ainmnt ur alve use dunn lant operations.
In e event valve leakage requires entry into Condition E, the Surveillance permits opening one valve in a penetration flow path to perform repairs. /N-627-SIR 3.6.3-2 This SR ensures that the Containment Air Release and Addition System isolation valves are closed as required or, if open, open for an allowable reason. If a valve is open in violation of this SR, the valve is considered inoperable.
If the inoperable valve is not otherwise known to have excessive leakage when closed, it is not considered to have leakage outside of limits. The SR is not required to be met when the valves are open for the reasons stated. The valves may be opened for pressure control, ALARA or air quality considerations for personnel entry, or for Surveillances that require the valves to be open. The valves are capable of closing in the environment following a LOCA. Therefore, these vpIves are allowed to be open for limited periods o~f time. -- hWl day ýF~ruency iscne io r containment iaion valv;,iequirem s -s S R This SR requires verification that each containment isolation manual valve and blind flange located outside containment or annulus and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. This SR does not require any testing or valve manipulation.
Rather, it involves verification, through system walkdown or computer status indication, that those containment isolation valves outside containment and capable of b i mis ositioned are in the correct position. "Sincev of valve, osition for containmeo"--'
isolaio a lves outsi lecontainment is r tively easy, the ency is base don engineering J,, me and was cho,,, n to provide j Sad tfd assuranc the correct p0 Ihe SR specifies- containment isolation va ves that are ope- lnder administrative controls are not required to meet the SR during th time the valves are open.Catawba Units 1 and 2 B 3.6.3-11 Revision Noo
ý'a CHA T Containment Isolation Valves HjR IOPATHIS0 0#4r;- B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)
This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be the correct position upon locking, sealing, or securing.The Note applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means 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 containment isolation valves, once they have been verified to be in the proper position, is small.SR 3.6.3.4 This SR requires verification that each containment isolation manual valve and blind flange located inside containment or annulus and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. For containment isolation valves inside containment, the Frequency of "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is appropriate since these containment isolation valves are operated under administrative controls and the probability of their misalignment is low. The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time they are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be the correct position upon locking, sealing, or securing.This Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification by administrative means 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 containment isolation valves, once they have been verified to be in their proper position, is small.Catawba Units 1 and 2 B 3.6.3-12 Revision No. 0 SCHANGES THIS PAGE. Containment Isolation Valves',1' CHAGES HIS AGE.B 3.6.3 FOR INFORMATION ONLY BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.3.5 Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY.
The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses.The isolation time is specified in the UFSAR and the Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.6.3.6 For the Containment Purge System valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J, Option B is required to ensure OPERABILITY.
The measured leakage rate for the Containment Purge System and Hydrogen Purge System valves must be < 0.05 La when pressurized to Pa.Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than other seal types. Based on this observation and the importance of maintaining this penetration leak tight (due to the direct path between containment and the environment), these valves will not be placed on the maximum extended test interval.
Therefore, these valves will be tested in accordance with Regulatory Guide 1.163, which allows a maximum test interval of 30 months.The Containment Air Release and Addition System valves have a demonstrated history of acceptable leakage. The measured leakage rate for containment air release and addition valves must be < 0.01 La when pressurized to Pa. These valves will be tested in accordance with Regulatory Guide 1.163, which allows a maximum test interval of 30 months.SR 3.6.3.7 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 Catawba Units 1 and 2 B 3.6.3-13 Revision No. 3 Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued) isolation signal. The isolation signals involved are Phase A, Phase B, and Safety Injection.
This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.
Th 18 month Frequency is bared on the need to perform this Survei Jaeftz under the conditions thatiapply during a plant outage and the potentifor an unplanned transienf the Surveillance
!were performed with/tie reactor at power. Oper ting experience has shown that these mponents usually pass thi Surveillance when performed at th 8 mon oFrequency.
Thefore, the Frequency was concluded to acceptable from a reliabili standpoint SR 3.6.3.8 This SR ensures that the combined leakage rate of all reactor building bypass leakage paths is less than or equal to the specified leakage rate.This provides assurance that the assumptions in the safety analysis are met. The Frequency is required by the Containment Leakage Rate Testing Program. This SR simply imposes additional acceptance criteria.Bypass leakage is considered part of La.REFERENCES
- 1. UFSAR, Section 15.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 3. UFSAR, Section 6.2.4. Standard Review Plan 6.2.4.5. Generic Issue B-24.Catawba Units 1 and 2 B 3.6.3-14 Revision Noo Containment Pressure B 3.6.4 BASES SURVEILLANCE REQUIREMENTS SR 3.6.4.1 Verifying that containment pressure is within limits ensures that unit operation remains within the limits assumed in the containment anal s.12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Fre ency of this SR developed based ondperating
" experience r eted to trending o fontainment pressure v iations during the appli 1Pol MODES. Furt! rmore, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Fr ztuency is consi red adequate in of other indications a ialable in the control ro, including alarmsjt alert the operator to afabnormal conta.inmentccrdtiý:r REFERENCES
- 1. UFSAR, Section 6.2.2. 10 CFR 50, Appendix K.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.6.4-4 Revision No-@
Containment Air Temperature B 3.6.5 BASES SURVEILLANCE REQUIREMENTS SR 3.6.5.1 and SR 3.6.5.2 Verifying that containment average air temperature is within the LCO limits ensures that containment operation remains within the limits assumed for the containment analyses.
In order to determine the containment average air temperature, an arithmetical average of ambient air temperature monitoring stations is calculated using measurements taken at locations within the containment selected to provide a representative sample of the overall containment atmosphere.
The upper compartment measurements should be taken at elevation 653 feet at the inlet of each operating upper containment ventilation unit. The lower compartment measurements should be taken at elevation 570 feet at the inlet of each operating lower containment ventilation unit.TIhe 24,our th-eseus is'considered ace~ptable based on o slow rates of tern jrature increase wit/yin containment as a/ suit of environmental Jat sources, (due to fe large volume of c eftainment).I Furthermore 4he 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequezj is considered ad, euate in view ofI ,other in qtions available in th econtrol room, inclu clg alarms, to alert the Oilrator to an abnormal temperure condition
.REFERENCES
- 1. UFSAR, Section 6.2.2. 10 CFR 50.49.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.6.5-4 Revision No.0 Containment Spray System B 3.6.6 BASES ACTIONS A.1 With one containment spray train inoperable, the affected train must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The components in this degraded condition are capable of providing 100% of the heat removal after an accident.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time was developed taking into account the redundant heat removal and iodine removal capabilities afforded by the OPERABLE train and the low probability of a DBA occurring during this period.B.1 and B.2 If the affected containment spray train 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 84ý hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. The extended interval to reach MODE 5 allows additional time and is reasonable when considering that the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.SURVEILLANCE SR 3.6.6.1 REQUIREMENTS Verifying the correct alignment of manual, power operated, and automatic-- valves, excluding check valves, in the Containment Spray System 67 provides assurance that the proper flow path exists for Containments c.ss "Spray System operation.
This SR does not apply to valves that are!locked, sealed, or otherwise secured in position since they were verified S ' 'in the correct position prior to being secured. This SR does not requireany testing or valve manipulation.
Rather, it involves verification, through a system walkdown or computer status indication, that those valves outside containment and capable of potentially being mispositioned, are in the correct position.
2 F C-? Ad SR 3.6.6.2 Verifying that each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head Catawba Units 1 and 2 B 3.6.6-5 Revision No.Y F6 Containment Spray System B 3.6.6 BASES SURVEILLANCE REQUIREMENTS (continued) ensures that spray pump performance has not degraded during the cycle.Flow and differential head are normal tests of centrifugal pump performance required by the ASME Code (Ref. 6). Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on bypass flow. This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.
The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.6.6.3 and SR 3.6.6.4 These SRs require verification that each automatic containment spray valve actuates to its correct position and each containment spray pump starts upon receipt of an actual or simulated Containment Phase B Isolation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative ofcoihe 18 s iontiFrequency is base~on the need SR 36. A se Surveillance under the conditios to si pspy during a plant/eutage and the pote ial for an unplanned, traietf the Sureillances were penrmed with the reactor arower. Operating perience has hor these components usi lyepass the Surveillances when performe the 18 month Frequen Therefore, the Symy was concludRIto be acceptable from a3po ine.The. surveillance of containment sumpisolation valves is also re SR 3.6.6.3. A single surveillance may be used to satisfy both requirements.
SIR 3.6.6.5 and SIR 3.6.6.6 These.SRs require verification of proper interaction between the CPCS system and the Containment Spray System.SR 3.6.6.5 deals solely with the containment spray pumps. It must be shown through testing that: (1) the containment spray pumps are prevented from starting in the absence of a CPCS permissive, (2) the containment spray pumps start when given a CPCS permissive, and (3)when running, the containment spray pumps stop when the CPCS permissive is removed. The "inhibit", "permit", and "terminate" parts of the CPCS interface with the containment spray pumps are verified by Catawba Units 1 and 2 B 3.6.6-6 Revision No.411 Containment Spray System B 3.6.6 BASES SURVEILLANCE REQUIREMENTS (continued) testing in this fashion.SR 3.6.6.6 deals solely with containment spray header containment isolation valves NS12B, NS15B, NS29A, and NS32A. It must be shown through testing that: (1) each valve closes when the CPCS permissive is removed, OR (2) each valve is prevented from opening in the absence of a CPCS permissive.
In addition to one of the above, it must also be shown that each valve opens when given a CPCS permissive.
SR 3.6.6.7 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections.
The spray nozzles can also be periodically tested using a vacuum blower to induce air flow through each nozzle to verify unobstructed flow. This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment during an accident is not degraded. 7e of thfe d-esign of the rlozzle, a te 10 yen,,tervals is £,,sidered adeqguat to detect obgfiruction of REFERENCES
- 1. 10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41, GDC 42, and GDC 43.2. UFSAR, Section 6.2.3. 10 CFR 50.49.4. 10 CFR 50, Appendix K.5. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 6. ASME Code for Operation and Maintenance of Nuclear Power Plants.Catawba Units 1 and 2 B 3.6.6-7 Revision No.U HSS B 3.6.8 BASES SURVEILLANCE SR 3.6.8.1 REQUIREMENTS Operating each HSS train for >_ 15 minutes ensures that each train is OPERABLE and that all associated controls are functioning properly.
It also ensures that blockage, fan and/or motor failure or excessive' Vibration detected fo)r cor-rective action. Plh e2 dyFre, ncy is " consisent with Ins ice ]est rng PulOga Su/ance Frgncies, I op exp. -nce, the kno (dfhe fan mQ r)s and controls, An_ the tw, na~in redunda
-SR 3.6.8.2 Verifying HSS fan motor current at rated speed with the motor operated suction valves closed is indicative of overall fan motor performance.
Since these fans are required to function during post-accident situations, the air density that the fans experience during surveillance testing will be different than the air density following a LOCA. An air density adjustment will be made to the average fan motor current test data before it is compared to the Technical Specification SR acceptance criteria.
Such inservice tests confirm component OPERABILITY, trend performance and detect incipient failures by indicating, abnormal performance.
("h-)Frequency,4f 92 days w based on eratn expernce whic J showns Frequenc4o be accep.le SR 3.6.8.3 This SR verifies the motor operated suction valves open upon receipt of a Containment Pressure -High High signal and associated time delay and that the HSS fans receive a start permissive when the valves start to opn e ec-y ot was -base~ln opnc fig Caawb U 1 ands 2 uencyBto3be.68-ta4 Revision Catawba Units 1 and 2 B 3.6.8-4 Revision No.{,9)
HSS B 3.6.8 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.8.4 This SR ensures that each HSS train responds properly to a containment pressure high-high actuation signal. The Surveillance verifies that each fan-starts after a delay of > 8 minutes and _ 10 minutes. Fre-que cy of 92 days coi orms with the te ing requirements r similar ESF equipment nd considers th nown reliability o an motors an controls and the o train redundd cy available.
Th fore, the Frep ency was i co ded to be acc 7able from a reliab" y standpoi nt REFERENCES
- 1. 10 CFR 50.44.2. 10 CFR 50, Appendix A, GDC 41, 42, and 43.3. Regulatory Guide 1.7, Revision 2.4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.6.8-5 Revision No. 8 HIS B 3.6.9 BASES ACTIONS (continued) length of time after the event that operator action would be required to prevent hydrogen accumulation from exceeding this limit, and the low probability of failure of the OPERABLE HIS train. Alternative Required Action A.2, by frequent surveillances, provides assurance that the OPERABLE train continues to be OPERABLE.B. 1 Condition B is one containment region with no OPERABLE hydrogen ignitor. Thus, while in Condition B, or in Conditions A and B simultaneously, there would always be ignition capability in the adjacent containment regions that would provide redundant capability by flame propagation to the region with no OPERABLE ignitors.Required Action B. 1 calls for the restoration of one hydrogen ignitor in each region to OPERABLE status within 7 days. The 7 day Completion Time is based on the same reasons given under Required Action A.1.C.1 The unit must be placed in a MODE in which the LCO does not apply if the HIS subsystem(s) cannot be restored to OPERABLE status within the associated Completion Time. This is done by placing the unit 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 />. 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, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE SR 3.6.9.1 REQUIREMENTS This SR confirms that > 34 of 35 hydrogen ignitors can be successfully energized in each train. The ignitors are simple resistance elements.Therefore, energizing provides assurance of OPERABILITY.
The allowance of one inoperable hydrogen ignitor is acceptable because, although one inoperable hydrogen ignitor in a region would compromise redundancy in that region, the containment regions are interconnected so*that ignition in one region would cause burning to progress to the others (i.e., there is overlap in each hydrogen ignitor's effectiveness between Catawb Unit a6requency,,2f 92-4yRevis'on, a e l hrogHo ting e Catawba Units I and 2 B 3.6.9-4 Revision No. 0-HIS B 3.6.9 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.9.2 This SR confirms that the two inoperable hydrogen ignitors allowed by SR 3.6.9.1 (ie., one in each train) are not in the same containmentrequency,,f'92 days the Frr~erlev Ct ý.w,c ovides the oy rn A more detailed functional test is performed(ve
~ to verify system OPERABILITY.
Each ignitor is visually examined to ensure that it is clean and that the electrical circuitry is energized.
All ignitors, including normally inaccessible ignitors, are visually checked for a glow to verify that they are energized.
Additionally, the surface temperature of each ignitor is measured in calm, nonturbulent atmospheric conditions to be> 1700'F to demonstrate that a temperature sufficient for ignition is achieved.
The 1700°F temperature is a surveillance requirement. "An Analysis of Hydrogen Control Measures at McGuire Nuclear Station" (Ref. 5) section 3.8 identifies that the required normal operation temperature is 1500'F. Therefore, based upon ignitor performance testing conducted at Catawba, the surveillance requirement of 1700'F ensures that sufficient margin is resent for continued hydrogen ignition under degraded bus conditions.2.
he 1 month Frequ is based on.the e1d tC orm 56 urveianca S der the ccnc s that apply(5.ing a Ant outagfand then potoial for an unpre ned transient if N furveill were/l erformed wi the reactor at .ower. OPerating,, exp~er" nce has own that t ise cornponen tus~ually pass the I-R when pe rmed at ine 18 mont.requenc, w is based fueling cCble. aherore, theUFrnitues w1 a 2 B 3o beviao le from a/alia bilitY 'a nd
~ REFERENCES
- 1. 10 CF.R 50.44.2. 10 CFR 50, Appendix A, GDC 41.3. UFSAR, Section 6.2.4. 10 CFR 50-86, Technical Specifications, (c)(2)(ii).
- 5. An Analysis of Hydrogen Control'Measures at.McGuire Nuclear Station-Catawba Units 1 and 2 B 3.6-9-5 R evisionNo AVS B 3.6.10 BASES ACTIONS (continued)
C.1 and C.2 If the AVS train 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.SURVEILLANCE SR 3.6.10.1 REQUIREMENTS Operating each AVS train from the control room with flow through the HEPA filters and carbon adsorbers ensures that all trains are OPERABLE and that all associated controls are functioning properly.
It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. Operation with the heaters on for _> 10 continuous hours eliminates moisture on the adsorbers and HEPA filters.Experience from filter testing at operating units indicates that the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> period is adequate for moisture elimination on the adsorbers and HEPA fitr.Te31 d g'requency was ve'loped in co~ederation of k Feliabi
- of fan motors d controls, thet,1o- train reduni_ ncy{available, ,fd the iodine r. oval capabilibty Qhe ConartSpy,.._Ice C.0nd j. it1 Ai 2...SR 3.6.10.2 This SR verifies that the required AVS filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The AVS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 5). The VFTP includes testing HEPA filter performance, carbon adsorber efficiency, system flow rate, and the physical properties of the activated carbon (general use and following specific operations).
Specific test frequencies and additional information are discussed in detail in the VFTP.Catawba Units 1 and 2 B 3.6.10-4 Revision No.
AVS'B 3.6.10 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.10.3 The automatic startup on ft injection signal ensures that each AVS train responds properly.
T-he vle th Fre ueno based a on tm need t fiter unit-that h ance undethe condgitionhat aay duD Lg a planMi lo oolingmay tbe netiay tor imittpeatue inreast if the Suieillance we performeuncith the reafr at power. filerating expadorence has a systemis verfied h the eablt of eachtritopdueheeqrd eown that thse compones usually pasTthe Surveeisas .c when r gurformed aethe 18 mont Frequency.
Th erefore the watequgecy a s pressuded be acceptole from a relioility standpoir.
Furt er ore, the SR eqif n -1 _ rn O P ER ILIT Y bI 6 s R 3a6310.4 The AVS filter cooling electric motor-operated bypass valves are tested to verify OPERABILITY.
The valves are normally closed and may need to be opened from the control room to initiate miniflow cooling through a filter unit that has been shutdown following a DBA LOCA. Miniflow cooling may be necessary to limit temperature increases in the idle filter tandue to decay heat from captured fission products..4"?he 18 Fep'cy is c s~idderee1" be acce ;ble baise~d/'n vallve rel/i' ility andýO g ,and ,te fact l t operatir
~ a h w tha h av s)t 'sua-- lly s. the rveil/ance, /en p .erf~or.ed at the l1 vtonth --SR 3.6..10.5 The proper functioning of the fans, dampers, filters, adsorbers, etc., as a system ~ _p _s v rfe byte ability of each train to produce the required ._system flow ae1Th.ef8 mon-fh Fr9eAuency with Reg e'ory
- f. e for~ntonaýtýsting. .SR 3.6.10.6 The ability of the AVS train to produce the required nega tive pressure of at least -0.88 inch water gauge when corrected to elevation 564 feet ensures that the annulus negative pressure is at least -0.25 inch water gauge everywhere in the annulus. The -0.88 inch water gauge annulus pressure includes a correction for an outs ide air temperature induced hydrostatic pressure gradient of -0.63 inch water gauge. The negative Catawba Units 1 and 2 B 3.6. 105 Revision No.
AVS B 3.6.10 BASES SURVEILLANCE REQUIREMENTS (continued) pressure prevents unfiltered leakage from the reactor building, since outside air will be drawn into the annulus by the negative pressure differential.
The CANVENT computer code is used to model the thermal effects of a LOCA on the annulus and the ability of the AVS to develop and maintain a negative pressure in the annulus after a design basis accident.
The annulus pressure drawdown time during normal plant conditions is not an input to any dose analyses.
Therefore, the annulus pressure drawdown time during normal plant conditions is insignificant.
The AVS trains are tested ro ensure each train will Lie- prat-i" -exper cls eacrailusull/~pasjthe.
rvei~~c w 06eipermed akte 18 c'h r, I r uenc uFurthermore, the SR interval was developed considering that_~ ~ ....
OPERABILITY is demonstrated at a 31 day Frequency by SR 3.6.10.1.
Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
REFERENCES
- 1. 10 CFR 50, Appendix A, GDC 41.2. UFSAR, Sections 6.2.3 and 9.4.9.3. UFSAR, Chapter 15.4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 5. Regulatory Guide 1.52, Revision 2.6. Catawba Nuclear Station License Amendments 90/84 for Units 1/2, August 23, 1991.7. NUREG-0800, Sections 6.2.3 and 6.5.3, Rev. 2, July 1981.Catawba Units 1 and 2 B 3.6.10-6'Revision No. 0 ARS B 3.6.11 BASES APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ARS. Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the ARS is not required to be OPERABLE in these MODES.ACTIONS A. 1 If one of the required trains of the ARS is inoperable, it must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time was developed taking into account the redundant flow of the OPERABLE ARS train and the low probability of a DBA occurring in this period.B.1 and B.2 If the ARS train 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.SURVEILLANCE SR 3.6.11.1 REQUIREMENTS Verifying that each ARS fan starts on an actual or simulated actuation signal, after a delay _> 8 minutes and _ 10 minutes, and operates for >_ 15 minutes is sufficient to ensure that all fans are OPERABLE and that all associated controls and time delays are functioning properly.
It also ensures that blockage, fan and/or motor failure, or excessive vibration cCatawba United for cand 2 B 31 Ray eviquency
./deveed co ~idering--e knnre biity of f~a, a lcontrols (,n et ain re h~d ancy a ;ible.I- i(FTL Catawba Units I and 2 B 3.6.11-4 Revision Noo ARS B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.11.2 Verifying ARS fan motor current at rated speed with the return air dampers closed confirms one operating condition of the fan. This test is indicative of overall fan motor performance.
Since these fans are required to function during post-accident situations, the air density that the fans experience during surveillance testing will be different than the air density following a LOCA. An air density adjustment will be made to the average fan motor current test data before it is compared to the Technical Specification SR acceptance criteria.
Such inservice tests confirm component OPERABILITY, .trend performace
_an frt incipient failures by indicating abnormal performane.
{The Frequ. e of-f92 days c rms with the reurmet' sýýlrES ýc Sequip :nt and considersAe known re~abaaili, yof fan controls, a he two train red,, dancy available INS ,RT , SR 3.6.11.3 Verifying the OPERABILITY of the return air damper provides assurance that the proper flow path will exist when the fan is started. This Surveillance also tests the circuitry, including time delays to ensure the sytem operates properly.
[T Frequen of 92 days d ev-elofp~e~d.,/
co0nsi ring t importan/
eof the da 5ers, theirlo ion, ph'y sic-en onm ,and pro ityffaleOperatin.experience sasa of. fail e. Op2! Ialso£is Frequ yto be acccctable.l..
SR 3.6.11.4 and SR 3.6.11.5 Verifying the OPERABILITY of the check damper in the air return fan discharge line to the containment lower compartment provides assurance that the proper flow path will exist when the fan is started and that flow can not occur when the fan is not operating.
Thjquncy nsidering the impo nce of {h~ pers, their locati, phsica/l vironment, and prob 'ility of faille Oerating experi e has also, own requenco be accce leJ FoC Wt e RF(s NA 'T,'Catawba Units 1 and 2 B 3.6.11-5 Revision Noo ARS B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.11.6 and SR 3.6.11.7 These SRs require verification that each ARS motor operated damper is allowed to open or is prevented from opening and each ARS fan is allowed to start or is de-energized or prevented from startilng based on the presence or absence of Containment Pressure Control System start permissive and terminate signals. The CPCS is described in the Bases for LCO 3.3.2, "ESFAS."The 18 m n Frquepcy is b on atj'expe45ncew has own it taebe accepj6elAe,, ,,- .,, REFERENCES
- 1. UFSAR, Section 6.2.2. 10 CFR 50, Appendix K.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 4. NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors." Catawba Units 1 and 2 B 3.6.11-6 Revision No.9 VLJ Ice Bed B 3.6.12 BASES ACTIONS A._1 If the ice bed is inoperable, it must be restored to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The Completion Time was developed based on operating experience, which confirms that due to the very large mass of stored ice, the parameters comprising OPERABILITY do not change appreciably in this time period. Because of this fact, the Surveillance Frequencies are long (months), except for the ice bed temperature, which is checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If a degraded condition is identified, even for temperature, with such a large mass of ice it is not possible for the degraded condition to significantly degrade further in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period.Therefore, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is a reasonable amount of time to correct a degraded condition before initiating a shutdown.B.1 and B.2 If the ice bed 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.SURVEILLANCE REQUIREMENTS SR 3.6.12.1 Verifying that the maximum temperature of the ic bed is < 27 0 F ensures that the ice is kept well below the melting point. The 12 h r Frequency was base n operatin experence, ich confirmed t t, due to the.large6i s of stored 'ce, it is not po ible for the i ed temperature to degra e significan within a 12 ur period and as also based on as ssing the pr ximity of the 0 limit to the elting temperatur urthermor ,the 12hour equency is c sidered adequat n view of indicatio in the contr room, incl e ar to al the o erator[to an nornb eThis SR may be satisfied by use of the Ice Bed Temperature Monitoring Stem.Catawba Units I and 2 B 3.6.12-5 Revision No-Z
- ,O CHANGES THIS PA,!. Ice Bed f -0FOIATl0A ONLY B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.2 This SR ensures that initial ice fill and any subsequent ice additions meet the boron concentration and pH requirements of SR 3.6.12.7.
The SR is modified by a NOTE that allows the chemical analysis to be performed on either the liquid or resulting ice of each sodium tetraborate solution prepared.
If ice is obtained from offsite sources, then chemical analysis data must be obtained for the ice supplied.SR 3.6.12.3 This SR ensures that the air/steam flow channels through the ice bed have not accumulated ice blockage that exceeds 15 percent of the total flow area through the ice bed region. The allowable 15 percent buildup of ice is based on the analysis of the sub-compartment response to a design basis LOCA with partial blockage of the ice condenser flow channels.
The analysis did. not perform detailed flow area modeling, but rather lumped the ice condenser bays into six sections ranging from 2.75 bays to 6.5 bays. Individual bays are acceptable with greater than 15 percent .blockage, as long as 15 percent blockage is not exceeded for any analysis section.To provide a 95 percent confidence that flow blockage does not exceed the allowed 15 percent, the visual inspection must be made for at least.54 (33 percent) of the 162 flow channels per ice condenser bay. The visual inspection of the ice bed flow channels is to inspect the flow area, by looking down from the top of the ice bed, and where view is achievable up from the bottom of the ice bed. Flow channels to be inspected are.determined by randorrrsample.,-As-the most restrictive ice bed flow passage is found at a lattice frame elevation, the 15 percent blockage criteria only applies to "flow channels" that comprise the area: a. between ice baskets, and b. past lattice frames and Wall panels.Due to a significantly larger flow area in the regions of the upper deck grating and the lower inlet plenum support structures and turning vanes, it would require a gross buildup of ice on these structures to obtain a degradation in air/steam flow. Therefore, these structures are excluded as part of a flow channel for application of the 15 percent blockage criteria.
Plant and industry experience have shown that removal of ice from the excluded structures during the refueling outage is sufficient to Catawba Units 1 and 2 B 3.6.12-6 Revision No. 4 Ice Bed B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued) ensure they remain operable throughout the operating cycle. Thus, removal of any gross ice buildup on the excluded structures is performed following outage maintenance activities.
Operating experience has demonstrated that the ice bed is the region that is the most flow restrictive, due to the normal presence of ice accumulation on lattice frames and wall panels. The flow area through the ice basket support platform is not a more restrictive flow area because it is easily accessible from the lower plenum and is maintained clear of ice accumulation.
There is not a mechanistically credible method for ice to accumulate on the ice basket support platform during plant operation.
Plant and industry experience has shown that the vertical flow area through the ice basket support platform remains clear of ice accumulation that could produce blockage.
Normally only a glaze may develop or exist on the ice basket support platform which is not significant to blockage of flow area. Additionally, outage maintenance practices provide measures to clear the ice basket support platform following maintenance activities of any accumulation of ice that could block flow areas.Activities that have a potential for significant degradation of flow channels should be limited to outage periods. Performance of this SR following completion of these activities assures the ice bed is in an acceptable FO JZ1- condition for the duration of the operating cycle.A clt, 0 Frost buildup or loose ice is not to be considered as flow channelblockage, whereas attached ice is considered blockage of a flow channel.g , IZ.3 Frost is the solid form of water that is loosely adherent, and can be brushed off with the open hand. , ...ake.~ R 3. 16.12.4 , 45, Ice mass determination methodology is designed to verify the total as-found (pre-maintenance) mass of ice in the ice bed, and the appropriateJ ,d fdistribution of that mass, using a random sampling of individual baskets., ; ,'The random sample will include at least 30 baskets from each of three d IEdefined Radial Zones (at least 90 baskets total). Radial Zone A consists ,_of baskets located in rowsý 8, add 9 (innermbst'rows adjacent to the 0+ rot..% / Crane Wall), Radial Zone B consists of baskets located in rows 4, 5, 6, ,,--^ ,;/ and 7 (middle rows of the ice bed), and Radial Zone C consists of baskets located in rows 1, 2, and 3 (outermost rows adjacent to the......._ :Containment Vessel). _ ...___ __........
Catawba Units 1 and 2 B 3.6.12-7 Revision No. 3 Ice Bed B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
The Radial Zones chosen include the row groupings nearest the inside and outside walls of the ice bed and the middle rows of the ice bed.w These groupings facilitate the statistical sampling plan by creating sub-populations of ice baskets that have similar mean mass and sublimation characteristics.
Methodology for determining sample ice basket mass will be either by direct lifting or by alternative techniques.
Any method chosen will include procedural allowances for the accuracy of the method used. The number of sample baskets in any Radial Zone may be increased once by adding 20 or more randomly selected baskets to verify the total mass of that Radial Zone.In the event the mass of a selected basket in a sample population (initial or expanded) cannot be determined by any available means (e.g., due to surface ice accumulation or obstruction), a randomly selected representative alternate basket may be used to replace the original selection in that sample population.
If employed, the representative alternate must meet the following criteria: a. Alternate selection must befrom the same bay-Zone (i.e., same bay, same Radial Zone) as the original selection, and b. Alternate selection cannot be a repeated selection (original or alternate) in the current Surveillance, and cannot have been used as an analyzed alternate selection in the three most recent Surveillances.
The complete basis for the methodology used in establishing the 95%confidence level in the total ice bed mass is documented in Ref. 5.The total ice mass and individual Radial Zone ice mass requirements defined in this Surveillance, and the minimum ice mass per basket requirement defined by SR 3.6.12.5, are the minimum requirements for OPERABILITY.
Additional ice mass beyond the SRs is maintained to address sublimation.
This sublimation allowance is generally applied to baskets in each Radial Zone, as appropriate, at the beginning of an operating cycle to ensure sufficient ice is available at the end of the operating cycle for the ice condenser to perform its intended design function.The equency 18 months w based on ice stoXge tests, and the t ical subli tion allowanc aintained in the* e mass over and ove the mini ice mass as med in the saftanalyses.
OperZiig and Catawba Units I and 2 B 3.6.12-8 Revision No.fO Ice Bed B 3.6.12 BAS ES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.5 Verifying that each selected sample basket from SR 3.6.12.4 contains at least 600 lbs of ice in the as-found (pre-maintenance) condition ensures that a significant localized degraded mass condition is avoided.4h>& r IZa-4a*This SR establishes a per basket limit to ensure any ice mass degradation is consistent with the initial conditions of the DBA by not significantly affecting the containment pressure response.
Ref. 5 provides insights through sensitivity runs that demonstrate that the containment peak pressure during a DBA is not significantly affected by the ice mass in a large localized region of baskets being degraded below the required safety analysis mean, when the Radial Zone and total ice mass requirements of SR 3.6.12.4 are satisfied.
Any basket identified as containing less than 600 lbs of ice requires appropriately entering the TS Required Action for an inoperable ice bed due to the potential that it may represent a significant condition adverse to quality.6 , 7\ re $ vee " c1t 0 TY As documented in Ref. 5, maintenance practices actively manage S__J individual ice basket mass above the required safety analysis mean for vv~o1 zed -l ..J5 .aS ach Radial Zone. Specifically, each basket is serviced to keep its ice9o~a~ £51 " ' lass above 750 lbs for Radial Zone A, 1196 lbs for Radial Zone B, and t..! ~I ) 96 lbs for Radial Zone C. If a basket sublimates below the safety fj, nalysis mean value, this instance is identified within the plant's.4 Ic .l action-pregram, includinglevaiuating-maintenance practices tou h. ' & dentify the cause and correct any deficiencies.
These maintenance 1' ,, 5 ractices provide defense in depth beyond compliance with the ice bed 5. urveillance requirements by limiting the occurrence of individual baskets S,~?~ ,, J& ith ice mass less than the required safety analysis mean.vA, % t r, SR 3.6.12.6IA5
?r , 1 ii, r his SR ensures that a representative sampling of accessible portions of{ f ke- I c b re f bbsRdS ,swh ici are relatively thin walled, perforated cylinders, have yr'~ -" r, e-, r, not been degraded by wear, cracks, corrosion, or other damage. The SR is designed around a full-length inspection of a sample of baskets, and is intended to monitor the effect of the ice condenser environment on ice baskets. The groupings defined in the SR (two baskets in each Catawba Units 1 and 2 B 3.6.12-9 Revision No.0 Ice Bed B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued) azimuthal third of the ice bed) ens ue that the sampling of baskets is resnbvditiue.The Frequency of 40 o0nths tor a vis al I (inspection o /he structural s'undne~ss of th/e ice baskets is ased on.Sengineeri judgment an ionsiders suc ,1actors as the icknesso basket relative to orrosion rates ,epected in th~f service(Senvir,,ment and the/, esults of the lop term ice st ~e ts1i~l SR 3.6.12.7 Verifying the chemical composition of the stored ice ensures that the stored ice has a boron concentration
> 1800 ppm and < 2330 ppm as sodium tetraborate and a high pH, _> 9.0 and < 9.5 at 25 0 C, in order to meet the requirement for borated water when the melted ice is used in the ECCS recirculation mode of operation.
Additionally, the minimum boron concentration setpoint is used to assure reactor subcriticality in a post LOCA environment, while the maximum boron concentration is used as the bounding value in the hot leg switchover timing calculation (Ref.4). This is accomplished by obtaining at least 24 ice samples. Each sample is taken approximately one foot from. the top of the ice of each randomly selected ice basket in each ice condenser bay. The SR is modified by a NOTE that allows the boron concentration and pH value obtained from averaging the individual samples' analysis results to satisfy the requirements of the SR. If either the average boron concentration or average pH value is outside their prescribed limit, then entry into ACTION Condition A is required.
Sodium tetraborate has been proven effective in maintaining the boron content for long storage periods, and it also enhances the ability of the solution to remove and retain fission product iodine. The high pH is required to enhance the effectiveness of the ice-- ,- and the melted ice in removing iodine froý i containment-atmosphere..-
This pH range also minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to ECCS and Containment S ray Sstem fluids in the recirculation mode of op:eration.
[The Frequency of 54 mo ths is intendedt withexpected le 'gth of three fuel c les, and was devel-ed cnsidering these facts: a. Lon/term ice storage t dsts have, determine hat the chemical /cc,onposition of the sted ice is extremelyable;
' /b. *'here are no nor (al operating, mech i~sms that significar y -change the bor concentration of t stored ice, and remains within a 9.0 -.5 range when bo concentrations e above approximat 1200 ppm; and Catawba Units 1 and 2 B 3.6.12-10 Revision No.f Ice Bed B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued) nce has d nstrated that m the boro pH r irements has no ,en a problem]REFERENCES
- 1. UFSAR, Section 6.2.2. 10 CFR 50, Appendix K.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 4. UFSAR, Section 6.3.3.5. Topical Report ICUG-001, Application of the Active Ice Mass Management Concept to the Ice Condenser Ice Mass Technical Specification, Revision 2.6. UFSAR, Section 18, Table 18-1.7. Catawba License Renewal Commitments, CNS-1274.00-00-0016, Section 4.17.Catawba Units 1 and 2 B 3.6.12-11 Revision No. 0 Ice Condenser Doors B 3.6.13 BASES ACTIONS (continued) 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 REQUIREMENTS SR 3.6.13.1 Verifying, by means of the Inlet Door Position Monitoring System, that the inlet doors are in their closed positions makes the operator aware of an inadvertent opening of one or more doors. JT-6 Freque~n t.,2hour_ensuresýare of thý,e eaust a e twh erd oor +k SR 3.6.13.2 Verifying, by visual inspection, that each intermediate deck door is closed and not impaired by ice, frost, or debris provides assurance that the intermediate deck doors (which form the floor of the upper plenum where frequent maintenance on the ice bed is performed) have not been left open or obstructed.
In determining if a door is impaired by ice, the frost accumulation on the doors, joints, and hinges are to be considered in--coniunction with the lifting force limits of SR 3.6.13.7.-
Frequen y of 7 dos is based n engineer g judgment .d takes it consider ion Ssh factors "the frequ cy of entry int,.the interm iate ice c ndenser eck the ti e required.or significant f st buildup, d the r that a DB will occu SR 3.6.13.3 Verifying, by visual inspection, that the top deck doors are in place and not obstructed provides assurance that the doors are performing their function of keeping warm air out of the ice condenser during normal operation, and wou'Ld not be obstructed if called upon to open in response to a DBA. [TVhFrequency of 9 days is based on ngineering judaftnent, wtconsi red such factorss the following:
/T/hrelative inaccessi ility and lack of t ffic in the vici of the/ors make it unlikel that a door wo be inadverte tlv left oD¢fn:i , Catawba Units 1 and 2 B 3.6.13-6/-ff Revision No.vi Ice Condenser Doors B 3.6.13 BASES SURVEILLANCE REQUIREMENTS (continued)
- c. The light nstruction of e doors wo d ensure th9k, in the eve6 e/ of a DA, air and gas s passing thr gh the ice o6ndenser uld find flow path, ev if a door we obstructed.
SR 3.6.13.4 Verifying, by visual inspection, that the ice condenser inlet doors are not impaired by ice, frost, or debris provides assurance that the doors a e free to o en in t event of a DBA. For this unitehe Freq-uenc f 18-mgs is base -n door desi~n, which doe not allow wa r Sco)ensation to eeze, and o/rating exper nce, which i icates th t M1e inlet do e rarely f to meet their R acceptan criteria.)
--Because of high radiation in the vicinity ot e inlet doors during power operation, this Surveillance is normally performed during a shutdown.SR 3.6.13.5 INSE AT Verifying the opening torque of the inlet doors provides assurance that no doors have become stuck in the closed position.
The value of 675* -lb Is based onn the vicin of the d oors during pb/ftr. operato r this Sureilancisnorall perfredsduri ng ah stdown o 10lb nit, toFrequency t fe i monthss based the passive natur h e the develop gmexcesism (ic tonae ret srgare no known ftors that Swuld change the ::ting, except buildup of ice; "e buildup dot likely, howev wt, because of the dtr design, whichni es not allowing: awater condenston to freeze). Op 2fating experiencvdicates thathe inlet doors .. ally meet their SPxscceptance crite .J Because of highiation in, te vicinity of the inlet cdoors duri-ng power" operation, this Surveillance is normally performed during a shutdown'
.-SR 3.6.13.6 The torque test Surveillance ensures that the inlet doors have not developed excessive friction and that the return springs are producing a*door return torque within limits. The torque test consists of the following:
Catawba Units 1 and 2 B 3.6.13-7 Revision No.0 Ice Condenser Doors B 3.6.13 BASES SURVEILLANCE REQUIREMENTS (continued)
- 1. Verify that the torque, T(OPEN), required to cause opening motion at the 400 open position is _< 195 in-lb;2. Verify that the torque, T(CLOSE), required to hold the door stationary (i.e., keep it from closing) at the 40° open position is >78 in-lb but < 250.6 in-lb; and 3. Calculate the frictional torque, T(FRICT) = 0.5 {T(OPEN) -T(CLOSE)}, and verify that the T(FRICT) is > -40 in-lb but < + 40 in-lb.-The purpose of the friction and return torque Specifications is to ensure that, in the event of a small break LOCA or SLB, all of the 24 door pairs open uniformly.
This assures that, during the initial blowdown phase, the steam and water mixture entering the lower compartment does not pass through part of the ice condenser leting the ice there, while bypassic g the ice in other bays. ih ra ini months is nsed Vn the rf sive OABILITY e inte rmediatse ite. d once ador p ed,ro t~here ,enoknowj factors that oul Ihn hestting, eept possibly, a buJ up of ice- fce buildup tlikl, h ever, of the door /d ign, whicl ~oes not allv wae o ston to fr ze). xseraencethth dicates tdc dors rare fe to oeni the e of ace ta.e cteria.fBecause of high radiation in the vicinity of the inlet doors during powetperation, this Surveillance is normally performed auring a shutdo ar.ri mo Verifying the OPERABILITY of the intermediate deck doors provides assurance that the intermediate deck doors are free to open in the event of a DBA. The verification consists of visually inspecting the intermediate doors for structural deterioration, verifying free movement of the vent assemblies, and ascertaining free movement of each door when'lifted with the applicable force shown below: Door Lifting Force a. Adjacent to crane wall 37.4 lb b. Paired with door adjacent to crane wall _ 33.8 lb c. Adjacent to containment wall _ 31.8 lb d. Paired with door adjacent to containment
< 31.0 lb wall Catawba Units 1 and 2 B 3.6.13-8 Revision No.-
Ice Condenser Doors B 3.6.13 BASES SURVEILLANCE REQUIREMENTS (continued)
T.he 18 nth Frequen is based on th passive desi of the interme iate deck do s, the frequenc of personnel try into the inter diate deck, d the fact that 3.6.13.2 co irms on a 7 da Fre uency that th doors are not i paired by ic , rost, or debri hich a ways a doo ould fail the oening force st (i.e., by stic g or from cnreased d weg REFERENCES
- 1. UFSAR, Chapter 6.2. 10 CFR 50, Appendix K.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 4. DPC-1201.17-00-0006, "Design and Licensing Basis for Ice Condenser Lower Inlet Doors Technical Specification Surveillance Requirements, 40° Opening, Closing and Frictional Torques." Catawba Units 1 and 2 B 3.6.13-9 Revision NO.0
"
PAGt. JDivider Barrier Integrity FOR6 INFORMATi.
N .ONLY B 3.6.14 BASES ACTIONS (continued) inspections in the pressurizer compartment during power operation and analysis performed that shows an open hatch (7.5 ft 2 bypass area) during a DBA does not impact the design pressure or temperature of the containment.
C.1 If the divider barrier seal is inoperable, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is allowed to restore the seal to OPERABLE status. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with LCO 3.6.1, which requires that containment 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 />.D.1 and D.2 If divider barrier integrity 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.SURVEILLANCE SR 3.6.14.1 REQUIREMENTS Verification, by visual inspection, that all personnel access doors and equipment hatches between the upper and lower containment compartments are closed provides assurance that divider barrier integrity is maintained prior to the reactor being taken from MODE 5 to MODE 4.This SR is necessary because many of the doors and hatches may have been opened for maintenance during the shutdown.SR 3.6.14.2 Verification, by visual inspection, that the personnel access door and equipment hatch seals, sealing surfaces, and alignments are acceptable provides assurance that divider barrier integrity is maintained.
This Catawba Units 1 and 2 B 3.6.14-4 Revision No. 0 Divider Barrier Integrity B 3.6.14 BASES SURVEILLANCE REQUIREMENTS (continued) inspection cannot be made when the door or hatch is closed. Therefore, SR 3.6.14.2 is required for each door or hatch that has been opened, prior to the final closure. Some doors and hatches may not be opened for longeriod of time. I 'ose that use/esilientimia~teria in the seals Ver 6-ifnbdy iu ins ectei aft nce everay 10opens to provide accss udora nthat the se material hat ot aged tb te hoint of degraded opertormance.
The F peqence of years is baseon the knownvd resiliency of e eals dr seals, the faec oration he Ctw Unt 1eneence taB6 confirms tl)tthe sea td at tis' Fr6quency Iave been found.,4 be/cceptable.iý---
SR 3.6.14.3 Verification, by visual inspection, after each opening of a personnel access door or equipment hatch that it has been closed makes the operator aware of the importance of closing it and thereby provides additional assurance that divider barrier integrity is maintained while in applicable MODES.SR 3.6.14.4 Conducting periodic physical property tests on divider barrier seal test coupons provides assurance that the seal material has not degraded in the containment environment, including the effects of irradiation with the reactor at power. -The required tests include a tensile strength test. Th-Fr uency of 1 months we"developd coidering such fctors as the/'i'lwn reiic f h ~ a era uscothe inacces lility of the s Is"and abse'e of traffic "their vicinitrz'nd the unit co l~itions needf to perform/ e S'R. Op eatin expeiee has show ffat these I com j;~ets usua 'Peln/
i i r"_~ ~~~uu pas ... t.. _ .. _ .---Catawbaabl Unit a rn liblt B 36.1-5eviionNo Divider Barrier Integrity B 3.6.14 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.14.5 Visual inspection of the seal around the perimeter ro that the seal is properly secured in place. T,1(e Frequency t'~s- veloped considring such factors the inaccessi and bsence of traff in their vicinity, th strength of th b rnechanisms used/to secure the seal, nd the unit co ditio ,erform the SR. perating experie ce has shown at thE components u ally pass the Su eillance when erformec 18 month Fr uency. Therefo , the Fre u as conc acceptable rom a reliability REFERENCES
- 1. UFSAR, Section 6.2.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.6.14-6 Revision No.10 Containment Recirculation Drains BASES B 3.6.15 SURVEILLANCE SR 3.6.15.1 and SR 3.6.15.2 REQUIREMENTS Verifying the OPERABILITY of the refueling canal drains ensures that they will be able to perform their functions in the event of a DBA. SR 3.6.15.1 confirms that the refueling canal drain valves have been locked open and that the drains are clear of any obstructions that could impair their functioning.
In addition to debris near the drains, SR 3.6.15.2 requires attention be given to any debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation and water is flowing to the drains. SR 3.6.15.1 must be performed before entering MODE 4 from MODE 5 after every filling of the canal to ensure that the valves have been locked open and that no debris that cou, impairthe drainsawas deposited durinn the time the canal was filled. SR 3.6A5.2 is performed ev 92 daysforthe uooDeq om artme t and refuel canal a as. 'The 92 Frequency s develop considering such ctors as the ccessibility of e drain Lthe ab f4'ratf. in M vicinity of tho06rains and th edunda of ZE J12 T2Z-SR 3.6.15.3 Verifying the OPERABILITY of the ice condenser floor drains ensures that they will be able to perform their functions in the event of a DBA.Inspecting the drain valve disk ensures that the valve is performing its function of sealing the drain line from warm air leakage into the ice condenser during normal operation, yet will open if melted ice fills the line following a DBA. Verifying that the drain lines are not obstructed ensures their, readiness to drain water from the ice cond~ennser.[-The 18 lFre develop, considerin-, such factors as'the inaccessi lity of tl dtrains durina ower operati n; the design of~the ice con riser, ,[ whn th'Sureilanceis/prfored t f_ 18monhe Freq Beas of high radiation in the vicinity of the drains during power oper tion, this Surveillance is normally done during a shutdown.REFERENCES
- 1. UFSAR, Section 6.2.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3- 15-4 Revision No.0 Reactor Building B 3.6.16 BASES APPLICABILITY Maintaining reactor building OPERABILITY prevents leakage of radioactive material from the reactor building.
Radioactive material may enter the reactor building from the containment following a LOCA.Therefore, reactor building OPERABILITY is required in MODES 1, 2, 3, and 4 when a LOCA or rod ejection accident could release radioactive material to the containment atmosphere.
In MODES 5 and 6, the probability and consequences of these events are low due to the Reactor Coolant System temperature and pressure limitations in these MODES. Therefore, reactor building OPERABILITY is not required in MODE 5 or 6.ACTIONS A.1 In the event reactor building OPERABILITY is not maintained, reactor building OPERABILITY must be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Twenty-four hours is a reasonable Completion Time considering the limited leakage design of containment and the low probability of a Design Basis Accident occurring during this time period.B.1 and B.2 If the reactor building 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.SURVEILLANCE REQUIREMENTS SR 3.6.16.1 Maintaining reactor building OPERABILITY requires maintaining the door in the access opening closed, except when the ac being used for normal transit ent and exit. The 31 daof1fequency iofs SR i~s hed Ton en~g~ee~naiedg if o is ne on er Uic g Jutdgme nd is contaered nai e u vieQw-o 466eotherjspdiGations of d tatus that areavailable.Jco Catawba Units 1 and 2 B 3.6.16-2 Revision No. z Reactor Building B 3.6.16 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.16.2 The annulus vacuum decay test is performed to verify the reactor building is OPERABLE.
A minimum annulus vacuum decay time of 87 seconds ensures that the reactor building design outside air inleakage rate is _< 2000,: cfm at an annulus differential pressure of -1.0 inch water gauge. Higher reactor building annulus outside air inleakage rates correlate to less holdup, mixing, and filtration of radiological effluents which increase offsite and operator doses.The vacuum decay test is performed by isolating the pressure transmitter and starting the AVS fan to draw down the annulus pressure to a significant vacuum. Isolating the transmitter enables the fan to reduce the annulus pressure below the normal setpoint.
The fan is then secured and the time it takes for the annulus pressure to decay or increase from-3.5 inches water gauge to -0.5 inch water gauge is measured.
The time required for the pressure in the annulus to increase from -3.5 inches water gauge to -0.5 inch water gauge is known as the vacuum decay time.The reactor building annulus outside air inleakage is an input to the CANVENT computer code, which provides input to the dose analyses.The CANVENT computer code is used to model the thermal effects of a LOCA on the annulus and the ability of the AVS to develop and maintain a negative pressure in the annulus after a design basis accident.
The code also determines AVS exhaust and recirculation airflow rates following a LOCA. The results of the CANVENT analysis for annulus conditions and AVS response to the LOCA also are used for the rod ejection accident.The 2000 cfm at -1.0 inch water gauge reactor building annulus outside air inleakage rate is conservatively corrected for ambient temperature and pressure as well as annulus differential pressure conditions prior to use as an input to the CANVENT computer code. The CANVENT results a-re then used as an input to the dose analyses.Z h co bi r-essuree b nary is te~sted~ery 18 montý. 1 ~month Fr ency is co tent with the g ance provideA NUREG Catawba Units 1 and 2 B 3.6.16-3 Revision Nof" Reactor Building B 3.6.16 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.16.3 ioration of the COl REFERENCES
- 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 2. UFSAR, Sections 6.2.3 and 6.2.6.5.3. NUREG-0800, Sections 6.2.3 and 6.5.3, Rev. 2, July 1981.Catawba Units 1 and 2 B 3.6.16-4 ReiinNo~
SG PORVs B 3.7.4 BASES SURVEILLANCE SR 3.7.4.1 REQUIREMENTS Verification of the nitrogen supply pressure on at least one tank for each SG PORV ensures that the SG PORVs will be available to mitigate the consequences of a steam enerator tube rupture concurrent with e loss of offsite ower. 'he 24 our frequdency is co)sistent~ith operg C-xpedr. e a p~has shotn to be Xcceptab! 'SR 3.7.4.2 To perform a controlled cooldown of the RCS, the SG PORVs must be able to be opened remotely and throttled through their full range using the safety-related nitrogen gas supply. This SR ensures that the SG PORVs are tested through a full control cycle at least once per fuel cycle.Performance of inservice testing or use of an SG PORV during a unit cooldown may satisfy this requirement.
0 eraing epeece hs- s wn--'thes componen~isUsually pa 'te Surveilla ce when perfo led at}the 1!'onthr n The Fquen~cy is a eptable from feliability/
SR 3.7.4.3 The function of the block valve is.to isolate a failed open SG PORV.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-unit cooldown may satisfy this requiremet pte -_-g e --ienceh s own-that lse compon ms'usualy t-he ---IqSurveillai perf 'ded at the 1 "onth Frequ h LFrequ cy is accept e from a rel*ility standlt.eft REFERENCES
- 1. UFSAR, Section 10.3.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.7.4-4 Revision Noe AFW System B 3.7.5 BASES ACTIONS (continued)
E.1 In MODE 4, either the reactor coolant pumps or the RHR loops can be used to provide forced circulation.
This is addressed in LCO 3.4.6, "RCS Loops-MODE 4." With one required AFW train inoperable, action must be taken to immediately restore the inoperable train to OPERABLE status. The immediate Completion Time is consistent with LCO 3.4.6.SURVEILLANCE REQUIREMENTS SR 3.7.5.1 Verifying the correct alignment for manual, power operated, and automatic valves in the AFW 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 that those valves capable of being mispositioned are in the correct position.
The SR is also modified by a note that excludes automatic valves when THERMAL POWER is< 10% RTP. Some automatic valves may be in a throttled position to support low power operation.
SR 3.7.5.2 Verifying that each AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME Code (Ref. 3). Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.
Performance of inservice testing discussed in the ASME Code (Ref. 3)(only required at 3 month intervals) satisfies this requirement.
Catawba Units 1 and 2 B 3-7.5-7 Revision No.0 AFW System B 3.7.5 BASES SURVEILLANCE REQUIREMENTS (continued)
This SR is modified by a Note indicating that the SR should be deferred until suitable test conditions are established.
This deferral is required because there is insufficient steam pressure to perform the test.SR 3.7.5.3 This SR verifies that AFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an ESFAS, 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 otheed in the .rqired osition unde administrative controls.The 18 m h Frequency i ased on the ne(to perform this Surveill ce under the c ditions that appl during a unit out e and the potent I for an unplan ed transient if th urveillance were erformed wit he reactor at p er. The 18 mo Frequency is a ptable b e o operating expe ence and the de gn reliability of t e ui ment This SR is modified by a Note that states the SR is not required in 4. In MODE 4, the required AFW train may already be aligned and operating.
SR 3.7.5.4 This SR verifies that the AFW pumps will start in the event of any accident or transient that generates an ESFAS by demonstrating that each AFW pump starts automatically on an actual or simulated actuation signal in MODES 1, 2, and 3. In MODE 4, the required pump may alreideady be operating and the autostart function is not required e18qunthFredIueO is4ba ed h te need to rq orm this Suldelancess ponvder the condimens thaf aopeo ating a uniuutagel and th tentiaore an unplanne if ransient Suop eillanc. ere erfor This SR is modified by two Notes. Note 1 indicates that the SR can be deferred until suitable test conditions are established.
This deferral is required because there is insufficient steam pressure to perform the test.Note 2 states that the SR is not required in MODE 4. In MODE 4, the required pump may already be operating and the autostart function is not required.
In MODE 4, the heat removal requirements would be less providing more time for operator action to manually start the required AFW pump if it were not in operation.
Catawba Units 1 and 2 B 3.7.5-8 Revision No.l CSS B 3.7.6 BASES ACTIONS (continued)
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 4, without reliance on the steam generator for heat removal, 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 unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.7.6.1 This SR verifies tl] the CSS contains the required invntr of wae.he 1-2 h rFeunyi b edon ope igepdter o er q uwaenes bs,/ref uievutin eqta may afe~c nt~heCS[ etoyIee hck/l=,te12 uh~ Frqecy is/osdrdviw of oth rindicatios in/~' controlI room/,nldn .aid. r ,toa let n thew of raort o anoralevations in cSn.Sl rom n leven ala toaetteo!ýtrt bora eitosi-eC.:
REFERENCES
- 1. UFSAR, Section 10.4.2. UFSAR, Chapter 6.3. UFSAR, Chapter 15.Catawba Units 1 and 2 B 3.7.6-3 Revision No. 0
"'0 C4 'NES TM~ PAC-.F 0 14 INF0RAA TIC10 :'l LV CCW System B 3.7.7 BASES APPLICABILITY In MODES 1, 2, 3, and 4, the CCW System is a normally operating system, which must be prepared to perform its post accident safety functions, primarily RCS heat removal, which is achieved by cooling the RHR heat exchanger.
In MODE 5 or 6, the requirements of the CCW System are determined by the systems it supports.ACTIONS A. 1 Required Action A. 1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," be entered if an inoperable CCW train results in an inoperable RHR loop.This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.
If one CCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, the remaining OPERABLE CCW train is adequate to perform the heat removal function.The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this period.B.1 and B.2 If the CCW train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit 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 unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.7.1 REQUIREMENTS 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 System.Verifying the correct alignment for manual, power operated, and automatic valves in the CCW flow path to safety related equipment provides assurance that the proper flow paths exist for CCW operation.
Catawba Units 1 and 2 B 3.7.7-3 Revision No. 0 CCW System B 3.7.7 BASES SURVEILLANCE REQUIREMENTS (continued)
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 that those valves capable of being mispositioned are in the correct position.Th 1day Fr uency is bas on engineer' judg~ment, is tsistent Swith the pedural cont oerning v fe operation, a ensures.j-core alve position SR 3.7.7.2 This SR verifies proper automatic operation of the CCW valves on an actual or simulated actuation safety injection, Phase 'A' Isolation, or Phase 'B' Isolation signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation.
This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required nosition under adiitaiecnrl.he 1 Freyen~cy is ba2'd on the needl to S vrmifis roeir ance oderthe tiono the W puy dsringan Stuage and thspotential ftoan unplanl d transienth O the Survei ncea oreperanorg ed withaheteactor at p uer. dper atng experine ehasi Cashaw Ui 1 and 2 B 7 paRevihiuoNeillanc whe F pe d tte18ot ree .fre. the Fz~quencyis SR 3.7.7.3 This SR verifies proper automatic operation of the CCW pumps on an actual or simulated actuation signal. The CCW System is a normally operating system that cannot be fully acutda art of ruietestinq during normal operainrhe 118- mnth Fr uency is based on t t need
... laegnce under the co 'itions that apply d gng a unit o ag~e and the po tial for an unpla ed transient if the S elac/wre performe lith the reactor at/l ower. Operating exr rience has /shon ha tsecomponents u :ally pass the Surveil ~lnce when pefomet he18 month Fr uency. T~herefore, tl Frequency is!".ac~acep l0e from a reliabili an " -i;t NS AT I Catawba Units 1 and 2 B 3.7.7-4 Revision No.(
NSWS B 3.7-8 BASES SURVEILLANCE SR 3.7.8.1 REQUIREMENTS This SR is modified by a Note indicating that the isolation of the NSWS components or systems may render those components inoperable, but does not affect the OPERABILITY of the NSWS.Verifying the correct alignment for manual, power operated, and automatic valves in the NSWS flow path provides assurance that the proper flow paths exist for NSWS 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 require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inauveriefive, as-check Xas ccalv..The 31 d Frequen is based on ngineering dgment, is nsiste Iwith th eproceduraJ ontos gnerinq vle 1er~ain an essu r es Cocrre tvalve po SR 3.7.8.2 This SR verifies proper automatic operation of the NSWS valves on an actual or simulated actuation signal. The signals that cause the actuation are from Safety Injection and Phase 'B' isolation.
The NSWS 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 th required position under administrative 18m mth Freque is
'need0 [o/: 1rformi i this TSurve nce i nsdefhe conoitie that appl uring a unitotage and the pmo tial for an uhplanned trminsient in thesurveillance tere perfoNme iy the reactoprat power. n perating eerience haihown that taent/'lmmPOnents dsually pas the Survei nce when e!rformed at Je 18 month Ffequec eeoreFequ is from a _rliahilit
_ _ -ttan d oi n .. ____This SR is modified by a Note that states that the SR is not required to be m~et for valves that are maintained in position to support NSWS single supply header operation.
When the NSWS is placed in this alignment, certain automatic valves in the system are maintained in position and will not automatically reposition in response to an actuation signal while the NSWS is in this alignment.
Catawba Units 1 and 2 B 3.7.8-6 Revision No.'
NSWS B 3.7.8 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.8.3 This SR verifies proper automatic operation of the NSWS pumps on an actual or simulated actuation signal. The signals that cause the actuation are from Safety Injection and Loss of Offsite Power. The NSWS is a normally operating system that cannot be fully actuated as part of normal tetndrng normal operation.
The 18 /zlonth Frequenc-y is ased on" ,11e e -erform thi sfurveillance u , er t he conditions/Clat apply d duringut outage the potentia for an unplanned ansient if the[Surv eflance were with t at pow e,. Operating/ee rence has s own that these domponents usu y pass the urveillance w n performed the 18 month Fr quencv. Th fore, the, Frequenc i cceptable fror a reliability sta point r -REFERENCES
- 1. UFSAR, Section 9.2-2. UFSAR, Section 6.2.3. UFSAR, Section 5.4.4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.7.'8-7 Revision Noo SNSWP B 3.7.9 BASES LCO The SNSWP is required to be OPERABLE and is considered OPERABLE if it contains a sufficient volume of water at or below the maximum temperature that would allow the NSWS to operate for at least 30 days following the design basis accident without the loss of net positive suction head (NPSH), and without exceeding the maximum design temperature of the equipment served by the NSWS. To meet this condition, the SNSWP temperature should not exceed 95°F at 568 ft mean sea level and the level should not fall below 571 ft mean sea level during normal unit operation.
APPLICABILITY In MODES 1, 2, 3, and 4, the SNSWP is required to support the OPERABILITY of the equipment serviced by the SNSWP and required to be OPERABLE in these MODES.In MODE 5 or 6, the requirements of the SNSWP are determined by the systems it supports.ACTIONS A.1 If the SNSWP is inoperable the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit 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 unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.9.1 REQUIREMENTS This SR verifies that adequate long term (30 day) cooling can be maintained.
The specified level also ensures that sufficient NPSH is avabl tooerate the NSWS pumpvalbetops.I The Frequenc>4-<based
[-on "o-perati gexperience/rated::to:--renoin9 ifthep aarame evariations
, drn eap licable DES his SR verifies that the SNSWP water level is 571 ft mean sea level.SR 3.7.9.2 This SR verifies that the NSWS is available to cool the CCW System to at least its maximum design temperature with the maximum accident or normal design heat loads for 30 days following a Design Basis Accident.Catawba Units 1 and 2 B 3.7.9-3 Revision No. (t)
SNSWP B 3.7.9 BASES SURVEILLANCE REQUIREMENTS (continued)
Lle 24 nour equency is b, ed on ope 'tng experien£ related to'rendinq he parametv ariations
- kfrinq the appli ble MODES. This SRverifies that the average water temperature of the SNSWVP is
- 957F.The SR is modified by a note that states the Surveillance is only required'to be performed during the months of July, August, and September.
During other month ambient temperature is below the surveillance lim it. .REFERENCES
- 1. UFSAR, Section 9.2.2. Regulatory Guide 1.27.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.7.9-4 Revision No.(1ndnmnt N~os. and 2ý28 CRAVS B 3.7.10.B A S E S .-----------------
--- ------- .........SURVEILLANCE SR 3-7.10.1 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 train once every month provides an adequate check of this system. Monthly heater operations dry out, any moisture accumulated in the carbon from humidity in the ambient air. Systems with heaters must be operated from the control room for > 10 continuous hours with the heaters energized and flow through the HEPA filters and carbon adsorbersThe 31 d reque',-- ed -Jýe q u ip m e t,-h t rd u n Cyi-SR 3.7.10.2 This SR verifies that the required CRAVS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The CRAVS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 5).The VFTP includes testing the performance of the HEPA filter and carbon adsorber efficiencies and the physical properties of the activated carbon.Specific test Frequencies and additional information are discussed in detail in the VFTP.SR 3.7.10.3 This SR verifies that each CRAVS train starts and operates on an actual.orSinlae acuVon eFequencyV of Xmonths is basde'd on Ar in o 'r..eratin~xein~
ani o ssýwith the typical SR 3.7.10.4 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 consequences is no more than 5 rem TEDE and the CRE occupants are protected from hazardous chemicals and smoke. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences.
When unfiltered air inleakage is greater than the assumed flow rate, Condition B Catawba Units 1 and 2 B 3.7:10-7 Revision CRAVS: = T2S PA .I B 3-7-10 BASES SURVEILLANCE REQUIREMENTS (continued) must be entered- Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for the occupants following an accident.
Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3 (Ref- 9), which endorses, with exceptions, NEI 99-03, Section 8-4 and Appendix F (Ref. 7). These compensatory measures may also be used as mitigating actions as required by Required Action B.2. Temporary analytical methods may also be used as compensatory measures to restore OPERABILITY (Ref.8). Options for restoring the CRE boundary to OPERABLE status include changing the licensing basis DBA consequence analysis, repairing the CRE boundary, or a combination of these actions. Depending upon the nature of the problem and the corrective action, a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status.REFERENCES
- 1. UFSAR, Section 6.4.2. UFSAR, Section 9.4.1.3. .UFSAR, Chapter 15.4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 5. Regulatory Guide 1.52, Rev. 2.6. Catawba Nuclear Station License Amendments 90/84 for Units 1/2, August 23, 1991.7. NEI 99-03, "Control Room Habitability Assessment", June 2001.8. Letter from Eric J. Leeds (NRC) to James W. Davis (NEI) dated January 30, 2004, "NEI Draft White Paper, Use of Generic Letter 91-18 Process and Alternative Source Terms in the Context of Control Room Habitability", (ADAMS Accession No.ML040300694).
- 9. Regulatory Guide 1.196, Rev. 1.Catawba Units 1 and 2 B 3.7.10-8 Revision No. 0 CRACWS B 3.7.11 BASES SURVEILLANCE REQUIREMENTS SR 3.7.11.1 This SR verifies that the heat removal capability of the system is s'to maintain the temperature in the control room.a or below 0F/2r-re-quejp'y is appropri'e since signifi nt degrada5n of-theslow and i.ot expectedcrý r this time p"riod.-REFERENCES
- 1. UFSAR, Section 9.4.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 3. 10 CFR 50.67, Accident source term.4. Regulatory Guide 1.183, Revision 0.Catawba Units 1 and 2 B 3.7.11-4 Revision No 19 4)
...ABFVES , ,B 3.7.12 BASES ACTIONS (continued) hazards such as radioactive contamination, toxic chemicals, smoke, temperature and relative humidity, and physical security.
Preplanned measures should be available to address these concerns for intentional and unintentional entry into the condition.
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is reasonable based on the low probability of a DBA occurring during this time period and the use of compensatory measures.
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> " Completion Time is a typically reasonable time to diagnose, plan and possibly repair, and test most problems with the ECCS pump rooms pressure boundary.C.1 and C.2 If the ABFVES train or ECCS pump rooms pressure boundary cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit 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 experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.D.1 and D.2 With one or more ABFVES heaters inoperable, the heater must be restored to OPERABLE status within 7 days. Alternatively, a report must be initiated per Specification 5.6.6, which details the reason for the heater's inoperability and the corrective action required to return the heater to OPERABLE status.The heaters do not affect OPERABILITY of the ABEVES filter trains because carbon adsorber efficiency testing is performed at 300C and 95% relative humidity.
The accident analysis shows that site boundary radiation doses are within 10 CFR 50.67 limits during a DBA LOCA under these conditions.
SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Systems should be checked periodically to ensure that they function properly.
As the environment and normal operating conditions on this system are not severe, testing each train once a month provides an adequate check on this system. Monthly heater operations dry out any moisture that may have accumulated in the carbon from humidity in the ambient air. Systems with heaters must be operated from the control room: _ 10 continuous hours with flow through the HEPA filters and Catawba Units 1 and 2 B 3.7.12-4 Revision No. 2 ABFVES B 3.7.12 BASES SURVEILLANCE REQUIREMENTS (continued) cabnasres and with the he-atrs energized.
The 31 y Frequency?
SR 3.7.12.2 This SR verifies that the required ABFVES testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The ABFVES filter tests are in accordance with Reference
- 5. The VFTP includes testing HEPA filter performance, carbon adsorbers efficiency, system flow rate, and the physical properties of the activated carbon (general use and following specific operations).
The system flow rate determination and in-place testing of the filter unit components is performed in the normal operating alignment with both trains in operation.
Flow through each filter unit in this alignment is approximately 30,000 cfm. The normal operating alignment has been chosen to minimize normal radiological protection concerns that occur when the system is operated in an abnormal alignment for an extended period of time.Operation of the system in other alignments may alter flow rates to the extent that the 30,000 cfm +10% specified in Technical Specification 5.5.11 will not be met. Flow rates outside the specified band under these operating alignments will not require the system to be considered inoperable.
Certain postulated failures and post accident recovery operational alignments may result in post accident system operation with only one train of ABFVES in a "normal" alignment.
Under these conditions system flow rate is expected to increase above the normal flow band specified in Technical Specification 5.5.11. An analysis has been performed which conservatively predicts the maximum flow rate under these conditions is approximately 37,000 cfm. 37,000 cfm corresponds to a face velocity of approximately 48 ft/min that is significantly more than the normal 40 ft/min velocity specified in ASTM D3803-1989 (Ref. 10). Therefore, the laboratory test of the carbon penetration is performed in accordance with ASTM D3803-1989 and Generic Letter 99-02 at a face velocity of 48 ft/min. These test results are to be adjusted for a 2.27 inch bed using the methodology presented in ASTM D3803-1989 pdor to comparing them to the Technical Specification 5.5.11 limit. Specific test Frequencies and additional information are discussed in detail in the VFTP.Catawba Units 1 and 2 B 3.7.12-5 Revision No.0 ABFVES B 3.7.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.12.3 This SR verifies that each ABFVES train starts and operates with flow through the HEPA filters and carbon adsorbers on an actual or simulated actuation signal. 8 13nth Freal c~ is cooisf.,nt 't For 5:F -c?' II This SR verifies the pressure boundary integrity of the ECCS pump rooms. The following rooms are considered to be ECCS pump rooms (with respect to the ABFVES): centrifugal charging pump rooms, safety injection pump rooms, residual heat removal pump rooms, and the containment spray pump rooms. Although the containment spray system is not normally considered an ECCS system, it is included in this ventilation boundary because of its accident mitigation function which requires the pumping of post accident containment sump fluid. The Elevation 522 pipe chase area is also maintained at a negative pressure by the ABFVES. Since the Elevation 543 and 560 mechanical penetration rooms communicate directly with the Elevation 522 pipe chase area, these penetration rooms are also maintained at a negative pressure by the ABFVES. The ability of the system to maintain the ECCS pump rooms at a negative pressure, with respect to potentially unfiltered adjacent areas, is periodically tested to verify proper functioning of the ABFVES. Upon receipt of a safety injection signal to initiate LOCA operation, the ABFVES is designed to maintain a slight negative pressure in the ECCS pump rooms, with respect to adjacent areas, to prevent unfiltered LEAKAGE. The ABFVES will continue to overate in this mode CatabtUitsaf Bec 3 -qnal is reset.R.sinreque of 18 mo¢ en ith the.Addance p~eided Sci .5.1 /Catawba Units I and 2 B 3.7.12-6 Revision FHVES B 3.7.13 BASES ACTIONS (continued)
With the movement of recently irradiated fuel in the fuel handling building, two trains of FHVES are required to be OPERABLE and one in operation.
The movement of recently irradiated fuel must be immediately suspended, if one or more trains of FHVES are inoperable or one is not in operation.
This does not preclude the movement of an irradiated fuel assembly to a safe position.
This action ensures that a fuel handling accident with unacceptable consequences could not occur.B.1 and B.2 With one or more FHVES heaters inoperable, the heater must be restored to OPERABLE status within 7 days. Alternatively, a report must be initiated per Specification 5.6.6, which details the reason for the heater's inoperability and the corrective action required to return the heater to OPERABLE status.The heaters do not affect OPERABILITY of the FHVES filter trains because carbon adsorber efficiency testing is performed at 30 0 C and 95% relative humidity.
The accident analysis shows that site boundary radiation doses are within 10 CFR 50.67 limits during a fuel handling accident under these conditions.
SURVEILLANCE SR 3.7.13.1 REQUIREMENTS With the FHVES train in service, a periodic monitoring of the system for proper operation should be checked on a routine basis to ensure that the system is functionilng roperly;, e 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Fequency is suffi -Ilt to fensure pro ;foperation throoohthe HEP ,,nd carbon filteree nd is Sbased o :v~e known re/la ity of the e jment~f.._., Systems should be checked periodically to ensure that they function*ho ronrlv. eniron tal an5: normal oper conditions o not seve C Uesting each traia every montBiesi n
this system. -Catawba Units I and 2 B 3.7.13-3 Revision No.0 FHVES B 3.7.13 BASES REFERENCES 1.2.3.4.5.UFSAR, Section 6.5.UFSAR, Section 9.4.UFSAR, Section 15.7.Regulatory Guide 1.25.10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 6. Not used.7. Regulatory Guide 1.52 (Rev. 2).8. ~~E~tO80, Sejc6I 1 11I 1 9. 10 CFR 50.67, Accident source term.10. Regulatory Guide 1.183 (Rev. 0).11. Catawba Nuclear Station License Amendments 90/84 for Units 1/2, August 23, 1991.Catawba Units 1 and 2 B 3.7.13-5 Revision No. 0 FHVES B 3.7.13 BASES SURVEILLANCE REQUIREMENTS (continued) r6 tl eli:
dries o tnyrmoisture carbon m humidity in the .ent air. ystems with heaters must beýý rme control room o ontinuous hours with flow through the .. A filters and carbon adsorbers and with the heaters er:ý.rjed431dFeue~ý!`
on thnw re y _SR 3.7.13.3 This SR verifies that the required FHVES testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The FHVES filter tests are in accordance with Regulatory Guide 1.52 (Ref. 7).The VFTP includes testing HEPA filter performance, carbon adsorber efficiency, system flow rate, and the physical properties of the activated carbon (general use and following specific operations).
Specific test frequencies and additional information are discussed in detail in the VFTP.SR 3.7.13.4 This SR verifies the integrity of the fuel building enclosure.
The ability of the system to maintain the fuel building at a negative pressure with respect to atmospheric pressure is periodically tested to verify proper function of the FHVES. During operation, the FHVES is designed to maintain a slight negative pressure in the fuel building, to prevent unfiltered LEAKAGE. The FHVES is designed to maintain < -0.25 inches water gau e with respect to atmospheric pressure at a flow rat f:!344crr h~e 18 mopis (onaW GRyc BASI443 cfm.sistexd an__ __.de
- N SR ..o Operating the FHVES filter bypass damper is necessary to ensure that the system functions properly.
The OPERABILITY of the FHV r bypass damper is verified if it can be manually closed. &ý -mirD Catawba Units 1 and 2 B 3.7.13-4 Revision No.
Spent Fuel Pool Water Level B 3.7.14 BASES LCO The spent fuel pool water level is required to be >_ 23 ft over the top of irradiated fuel assemblies seated in the storage racks. The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 3). As such, it is the minimum required for fuel storage and movement within the spent fuel pool.APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel pool, since the potential for a release of fission products exists.ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring.
When the spent fuel pool water level is lower, than the required level, the movement of irradiated fuel assemblies in the spent fuel pool is immediately suspended to a safe position.
This action effectively precludes the occurrence of a fuel handling accident.
This does not preclude movement of a fuel assembly to a safe position.If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. 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 3.7.14.1 REQUIREMENTS This SR verifies sufficient spent fuel pool water is available in the event of a fuel handling accident.
The water level in the spent fuel pool must be The 74 Frequency is ap riate because t f-olume pool isnormy stable. Water,/el changes are c trolled byUplarIjafrocedures a are acceptable ed oe During refueling operations, the level in the spent fuel pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.Catawba Units 1 and 2 B 3.7.14-2 -Revision Not Spent Fuel Pool Boron Concentration B 3.7.15 BASES APPLICABLE SAFETY ANALYSES (continued)
The concentration of dissolved boron in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 7).LCO The spent fuel pool boron concentration is required to be within the limits specified in the COLR. The specified concentration of dissolved boron in the spent fuel pool preserves the assumptions used in the analyses of the potential critical accident scenarios as described in Reference
- 6. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the spent fuel pool.APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.When the concentration of boron in the fuel storage pool 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 concentration of boron is restored simultaneously with suspending movement of fuel assemblies.
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 3.7.15.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the anal zed accidents areý ful addressed day Fi'quency Fs a ;ro0priate,n 4tajor repleni me 01atris exPeoed to tak place_)over suc a short perio df Catawba Units 1 and 2 B 3.7.15-3 Revision No.)
Secondary Specific Activity B 3.7.17 BASES 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 RCSand contributes to increased post accident doses, If the secondary specific activity cannot be restored to within limits within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit 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 unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.7.17.1 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.LEAKAG h~e 31 ency is based 0n9t-8 detecti~r' increasing ends of the level o SE EQUIVALENT-131, and ows for iate action to b aken to maintain levei's below thertco Cjmif REFERENCES
- 1. 10 CFR 50.67.2. UFSAR, Section 15.1.5.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 4. Regulatory Guide 1.183, July 2000.Catawba Units 1 and 2 B 3.7.17-3 Revision No AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.1 This SR ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure that distribution buses and loads are connected to their preferred power source, and r riat independence of off site circuits is maintained.1Th'd a y; _lr eý. ncy i s J1~ si_ e breaker estio es~ leyýroange wit pout theoperat r'ei n~wared5f it and i~cue it stt~ sds~~en th~e SR 3.8.1.2 and SR 3.8.1.7 These SRs help to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition.
To minimize the wear on moving parts that do not get lubricated when the engine is not running, these SRs are modified by a Note (Note 2 for SR 3.8.1.2) to indicate that all DG starts for these Surveillances may be preceded by an engine prelube period and followed by a warmup period prior to loading.For the purposes of SR 3.8.1.2 and SR 3.8.1.7 testing, the DGs are started from standby condifions using a manual start, loss of offsite power signal, safety injection signal, or loss of offsite power coincident with a safety injection signal. Standby conditions for a DG mean that the diesel engine coolant and oil are being continuously circulated and temperature is being maintained consistent with manufacturer recommendations.
In order to reduce stress and wear on diesel engines, the manufacturer recommends a modified start in which the starting speed of DGs is limited, warmup is limited to this lower speed, and the DGs are gradually accelerated to synchronous speed prior to loading. These start procedures are the intent of Note 3, which is only applicable when such modified start procedures are recommended by the manufacturer.
Catawba Units 1 and 2 B 3.8.1-15 Revision -N AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.7 requires that, at 84 d Fre nc he DG starts from standby conditions and ach"eves required voltage and frequency within 11 seconds. The 11 second start requirement supports the assumptions of the design basis LOCA analysis in the UFSAR, Chapter 15 (Ref. 5).The 11 second start requirement is not applicable to SR 3.8.1.2 (see Note 3) when a modified start procedure as described above is used. If a modified start is not used, the 11 second start requirement of SR 3.8.1.7 applies.Since SR 3.8.1.7 requires a 11 second start, it is more restrictive than SR 3.8.1.2, and it may be performed in lieu of SR 3.8.1.2. This is the intent of Note 1 of SR 3.8.1.2.* e nor I31 day quency for S,.8.1.2 is consis t with fegul ory Guide .9 (Ref. 3). T 184 day Freq cy for S .Sa ruction in cfd testing con stent with Generc Letter -15 T ese Frequ ncies provide dequate assur ce of PERABILITY, while mini izing degrad ion resulting f testin SR 3.8.1.3 This Surveillance verifies that the OGs 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, while minimizing the time that the DG is connected to the offsite source.Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0.The 0.8 value is the design rating of the machine, while the 1.0 is an operational limitation to ensure circulating currents are minimized.
The load band is provided to avoid routine overloading of the DG. Routine overloading may.result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY.
Te31 e~lre._qurv ye" ShulrIeilla is consistentwKt This SR is modified by four Notes. Note 1 indicates that diesel engine runs for this Surveillance may include gradual Ioading,asrecommended Catawba Units 1 and 2 B 3.8.1-16 Revision Nor AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized.
Note 2 states that momentary transients, because of changing bus loads, do not invalidate this test. Similarly, momentary power factor transients above the limit do not invalidate the test. Note 3 indicates that this Surveillance should 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 DG start must precede this test to credit satisfactory performance.
SR 3.8.1.4 This SR provides verification that the level of fuel oil in the day tank is at or above the level at which fuel oil is automatically added. The level is expressed as an equivalent volume in gallons, .and is selected to ensure 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 full load plus 10%.T~~~re s dequate to htasficient s~u~o-,,fel oil is a able, since low level oarms are provided aýCilityr would be aware of ar*,1arge uses of fuel okbduring this period.S R 3.8.1.5 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environnent i to survive. Removal of water from the fuel oil day tanks4Cnc
- /3e a eliminates the necessary environment for bacterial survival.
This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation.
Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and breakdown of the fuel oil by bacteria.
Frequent checking for and removal of accumulated water minimizes fouling anA hrovides data re ardin the watertight integrity of.Re ulato.DP.-'uide 11 Ihis SR is for preventative maintenance.
The presence o water doe not necessarily represent failure of this SR, provided the accumulat d water is removed during the performance of this Surveillance.
MEAT5~ I of r-"31£y Catawba Units 1 and 2 B 3.8.1-17 Revision AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.6 This Surveillance demonstrates that each required fuel oil system operates and transfers fuel oil from its associated storage tanks to its associated day tank. This is required to support continuous operation of standby power sources. This Surveillance provides assurance that the fuel oil valve is OPERABLE, and allows gravity feed of fuel oil to the day tank from underground storage, tanks, to ensure the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for fuel transfer systems are OPERABLE.Th einof f~uel nsfer sys-tems is at the transfer .operates au atically or the tra qsr valve bypass valve be opened manuall _ order to maintaina adequate volume of f oil in the day tank urin r fotlowin testinq. Therefore, ,-TI day Frequen propriate.
SR 3.8.1.7 See SR 3.8.1.2.SR 3.8.1.8 Transfer of each 4.16 kV ESE bus power supply from the normal offsite circuit to the alternate offsite circuit demonstrates the capability of the* alternate circuit distribution network to power the shutdown loads. The alternate circuit distribution network consists of an offsite power source through a 6.9 kV bus incoming breaker, its associated 6.9 kV bus tie breaker and the aligned 6.9/4.16 kV transformer to the essential bus.The requirement of this SR is the transfer from the normal offsite circuit to the alternate offsite circuit via the automatic and manual actuation of the 6.9 kV bus tie breaker and 6.9 kV bus incoming breakers upon loss of the normal offsite source that is being credited.
Capability of manualla s appinin to,a standby tran-sformer is nt required to satisfy this SR .The'-1ý81mont Surveillanc~i-.based otngineering
-judg_ m t, taking in t/conside ratio n t unit condIti fs required to pe rm the Surve, ance, and is int ded to be nsistent with e ected f el cycle length/._
Operating exprience has,;, own that these components U~ally pass the S when perf rned at the 18/ fonth .?Frequency.
/Therefore, the ab4_unc, e.._'cnld~~nI c from a relV ility stand oin.Catawba Units 1 and 2 B 3.8.1-18 Revision Nc AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.9 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 Surveillance demonstrates the DG load response characteristics and capability to reject the largest single load without exceeding predetermined voltage and. frequency and while maintaining a specified margin to the overspeed trip. For this unit, the single load for each DG and its horsepower rating is as follows: Nuclear Service Water pump which is a 1000 H.P. motor. This Surveillance may be accomplished by: a. Tripping the DG output breaker with the DG carrying greater than or equal to its associated single largest post-accident load while paralleled to offsite power, or while solely supplying the bus; or b. Tripping its associated single largest post-accident load with the DG solely supplying the bus.As required by Regulatory Guide 1.9 (Ref. 3), the load rejection test is acceptable if the increase in diesel speed does not exceed 75%of the difference between synchronous speed and the overspeed trip setpoint.The value of 63 Hz has been selected for the frequency limit for the load rejection and it is a more conservative limit than required by Reference 3.The time, voltage, and frequency tolerances specified in this SR are derived from Regulatory Guide 1.9 (Ref. 3) recommendations for response during load sequence intervals.
The 3 seconds specified is equal to 60% of a typical 5 second load sequence interval associated with sequencing of the largest load. The voltage and frequency specified are consistent -with the design range of the equipment powered by the DG. SR.3.8.1.9.a corresponds to the maximum frequency excursion, while SR 3.8.1.9.b and SR 3.8.1.9.c are steady state voltage and frequency value' o which the s stem must reco ing load rejection.
T 18 nth Fre ency is nsi nt This SR is modified by a Note. In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, the Note requires that, if synchronized to offsite power, testing must be performed using a power factor < 0.9. This power factor is chosen to be representative of the actual design basis inductive loading-_that--the-DG
,-would experience:--
Catawba Units 1 and 2 B 3.8-1-19 Revision Non, AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.10 This Surveillance demonstrates the DG capability to reject a full load without overspeed tripping or exceeding the predetermined voltage limits.The DG full load rejection may occur because of a system fault or inadvertent breaker tripping.
This Surveillance ensures proper engine generator load response under the simulated test conditions.
This test simulates the loss of the total connected load that the DG experiences following a full load rejection and verifies that the DG does not trip upon loss of the load. These acceptance criteria provide for DG damage protection.
While the DG is not expected to experience this transient during an event-and continues to be available, this response ensures that the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be corrected or isolated.Although not representative of the design basis inductive loading that the DG would experience, a power factor of approximately unity (1.0) is used for testing. This power factor is chosen in accordance with manufacturer's recommendations to minimize DG overvoltage damage during testing.T 18 mon requenc consistent with -recommend n of egulat Guide 1.1ý0 and is tended to be ~dfisistent ex ed fe y nt ,SR 3.8.1.11 As required by Regulatory Guide 1.108 (Ref. 10), paragraph 2.a.(1), this Surveillance demonstrates the as designed operation of the standby power sources during loss of the offsite source. This test verifies all actions encountered from the loss of off site power, including shedding of the nonessential loads and energization of the emergency buses and respective loads from the DG. It further demonstrates the capability of the DG to automatically achieve the required voltage and frequency within the specified time.The DG autostart time of 11 seconds is derived from requirements of the accident analysis to respond to a design basis large break LOCA. The Surveillance should be continued for a minimum of 5 minutes in order to demonstrate that all starting transients have decayed and stability is achieved.Catawba Units 1 and 2 B 3.8.1-20' Revision No.v AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The requirement to verify the connection and power supply of the emergency bus and autoconnected loads is intended to satisfactorily show the relationship of these loads to the DG loading logic. In certain circumstances,'
many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation.
For instance, Emergency Core Cooling Systems (ECCS) injection valves are not desired to be stroked open, or high pressure injection systems are not capable of being operated at full flow, or residual heat removal (RHR)systems performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation.
In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG systems 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.The Frequency 18 months ise.onsistent with the commendatio of Regulatory ide 1.108( paragraph 2 .(1), takes into )consider ron unit conditi s required to pe rm the Survei e, a -i inten to be consis nt with expecte el cycle lengths. /N$EeT ,2.This SR is modified by two Notes. The reason for Note 1 is 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 coola'nt and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
The reason for Note 2 is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.SR 3.8.1.12 This Surveillance demonstrates that the DG automatically starts and achieves the required voltage and frequency within the specified time (11 seconds) from the design basis actuation signal (LOCA signal) and operates for >- 5 minutes. The 5 minute period provides sufficient time to demonstrate stability.
SR 3.8.1.12.d ensures that the emergency bus remains energized from the off site electrical power system on an ESF signal without loss of offsite power.Catawba Units 1 and 2 B 3.8.1-21 Revision No.{0 AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
T Freque y of 18 months.?es into consid ýPation unit conditi s equired perform the Su eillance and is) ended to be constent with the ex cted fuel cycle I gths. Operati experience has own that thes components u ally pass the V when performed t the 18 mo h Fr quency. Theref e, the Fri ecu v was cocludeo t le om a reliability andpoirkhis SR is modified by a Note. The reason for the Note is 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 coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
SR 3.8.1.13 This Surveillance demonstrates that DG non-emergency protective functions (e.g., high jacket water temperature) are bypassed on a loss of voltage signal concurrent with an ESF actuation test signal. Non-emergency automatic trips are all automatic trips except: a. Engine overspeed;
- b. Generator differential current;c. Low -low lube oil pressure; and d. Voltage control overcurrent relay scheme.The non-emergency trips are bypassed during DBAs and 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. Currently, DG emergency automatic trips are tested periodically per the station periodic maintenance program.The 18 nth Frequ cy is based engineering~
dgment, taking i consi ration onditions re ired to perfor the Surveillancand is inte ded to b onsistent wi expected fuenycle lengths. erating perienc as shown th these compo nts usually pa the SR wh prform at the 18 nnth Frequenc I. Therfet euny s Y" eef t reque~ncy conc ded to be ac *ptable J ~maelaiiysa it SR 3.8.1.14 Regulatory Guide 1.108 (Ref. 10), paragraph 2.a.(3), requires demonstratioK(ho~
1 Znththat the DGs can start and run Catawba Units 1 and 2 B 3.8.1-22 Revision No.4 AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) continuously at full load capability for an interval of not less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The DG starts for this Surveillance can be performed either from standby or hot conditions.
The provisions for prelubricating and warmup, discussed in SR 3.8.1.2, and for gradual loading, discussed in SR 3.8.1.3, are applicable to this SR.In order to ensure that the DG is tested under load conditions that are as close to design conditions as possible, testing must be performed using a power factor of <O.97. This power factor is chosen to be representative of the actual design basis inductive loading that the DG would experience.
The load band is provided to avoid routine overloading of the DG.Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY.
requenc~yis carssent w' h~e recoo endations of.Hi Huaoruide 1.1080 ý a graph 2.a. , takes into , CISon s id ~ton unit con dt'nsrq*t rpero te Sreil a3[.ine .ed to be conp~tent with hhoecedfuercl This Surveillance is modified by a Note. The Note states that momentary transients due to changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the power factor limit will not invalidate the test.SR 3.8.1.15 This Surveillance demonstrates that the diesel engine can restart from a hot condition, such as subsequent to shutdown from normal Surveillances, and achieve the required voltage and frequency within 11 seconds. The 11 second time is derived from the requirements of the'dnt anlssto.r espond to desg bai !ag rak LOCA.TT;This SR is modified by two Notes. Note 1 ensures that the test is performed with the diesel sufficiently hot. The load band is provided to Catawba Units 1 and 2 B 3.8.1-23 Revision No.9 AC Sources--Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) avoid routine overloading of the DG. Routine overloads may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY.
The requirement that the diesel has operated for at least an hour at full load conditions prior to performance of this Surveillance is based on manufacturer recommendations for achieving hot conditions.
Momentary transients due to changing bus loads do not invalidate this test. Note 2 allows all DG starts to be preceded by an engine prelube period to minimize wear and tear on the diesel during testing.SR 3.8.1.16 As required by Regulatory Guide 1.108 (Ref. 10), paragraph 2.a.(6), this Surveillance ensures that the manual synchronization and automatic load-transfer from the DG to the offsite source can be made and the DG can be returned to standby operation when offsite power is restored.
It also ensures that the autostart logic is reset to allow the DO to reload if a subsequent loss of offsite power occurs. The DG is considered to be in standby operation when the DG is at rated speed and voltage, the output breaker is open and can receive an autoclose signal on bus undervoltage, and the load sequence timers are reset.Threu(y of onts, onsi nt with tre recorm midati st -gltKG ide X. 18 .__ ,1 rag ra pJ;2,a. (6), an ake " ,consic ration u ' conditi 04s re/ ired o rm th rvei. -This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.SR 3.8.1.17 Demonstration of the test mode override ensures that the DG availability under accident conditions will not be compromised as the result of testing and the DG will automatically reset to standby operation if a LOCA actuation signal is received during operation in the test mode. Standby operation is defined as the DG running at rated speed and voltage with the DG output breaker open. These provisions for automatic switchover are required by Regulatory Guide 1.9 (Ref. 3).Catawba Units 1 and 2 B 3.8.1-24 Revision No."
AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The requirement to automatically energize the emergency loads with off site power is essentially identical to that of SR 3.8.1.12.
The intent in the requirement associated with SR 3.8.1.17.b is to show that the emergency loading was not affected by the DG operation in test mode.In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the emergency loads 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.The month Fre cy is consistent with th& Iecommen aons of/,egulatory Gui1108 "(Ref. 10), para"h into consider n unit conditions requpie to perform veillarce, and is , inte d to be consistent with epected fuel cy e leng~th This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.SR 3.8.1.18 Under accident and loss of offsite power conditions loads are sequentially connected to the bus by the automatic load sequencer.
The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs due to high motor starting currents.
The load sequence time interval tolerance in Table 8-6 Df Reference 2 ensures that sufficient time exists for the DG to restore frequency and voltage prior to applying the next load and that safety analysis assumptions regarding ESF equipment time delays are not violated.Table 8-6 of Reference 2 provides a summary of the automatic loading of ESF buses.Frequenf of 18 months isp.ssistent with the. ns of Regulato Guide 1.108 (Re.j10), paragraph
.(2), take to consi ration unit cond ins required to orm the S eillace, Zndis lded to be consiatent with expected uel cycle I gth Catawba Units 1 and 2 B 3.8.1-25 Revision AC Sources-Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.19 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 Surveillance demonstrates the DG operation, as discussed in the Bases for SR 3.8.1.11, during a loss of offsite power actuation test signal in conjunction with an ESF 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.T e Freq cy of 18 ths takes int onsideratio t conditi requir o perform e Surveillanc and is int o be co tent with a ppecteed fue ycle length o 8 month'-This SR is modified by two Notes. The reason for Note 1 is 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 coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations for DGs. The reason for Note 2 is that the performance of the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.SR 3.8.1.20 This Surveillance demonstrates that the DG starting independence has not been compromised.
Also, this Surveillance demonstrates that each engine can achieve proper speed within the specified time when the DGs are started simultaneously.
C ga G de ,0 This SR is modified by a Note. The reason for the Note is to minimize wear on the DG during testing. For the purpose of this testing, the DGs Catawba Units 1 and 2 B 3.8.1-26 Revision No%
Diesel Fuel Oil, Lube Oil and Starting Air B 3.8.3 BASES ACTIONS (continued) for reasons other than addressed by Conditions A through E, the associated DG may be incapable of performing its intended function and must be immediately declared inoperable.
SURVEILLANCE REQUIREMENTS SR 3.8.3.1 This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for 7 days at full load. The 7 day period is sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.SR 3.8.3.2 This Surveillance ensures that sufficient lube oil inventory is available to support at least 7 days of full load operation for each DG. The 400 gal requirement is based on the DG manufacturer consumption values for the run time of the DG. In order to account for the lube oil sump tank inventory decrease that occurs when the DG is started, the 400 gal requirement shall be met with the Surveillance conducted while the DG is running.SR 3.8.3.3 The tests listed below are a means of determining whether new fuel oil is of the appropriate grade and has not been contaminated with substances that Would have an immediate, detrimental impact on diesel engine combustion.
If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks. These Catawba Units 1 and 2 B 3.8.3-5 Revision Catawba Units I and 2 B 3.8.3-5 Revision No 1,3 CHANGES THIS PAGE. Diesel Fuel Oil, Lube Oil and Starting Air-1C0R Imi"OR'ITl ON ONLY B 3.8.3 BASES SURVEILLANCE REQUIREMENTS (continued) tests are to be conducted prior to adding the new fuel to the storage tank(s). The tests, limits, and applicable ASTM Standards are as follows: a. Sample the new fuel oil in accordance with ASTM D4057 (Ref. 7);b. Verify in ýccordance with the tests specified in ASTM D975 (Ref. 7)*that the sample has a kinematic viscosity at 400C of > 1.9 centistokes and 4.1 centistokes, and a flash point of > 1250F; and c. Verify that the new fuel oil has a clear and bright appearance with proper color when tested in accordance with ASTM D4176 (Ref. 7) or a water and sediment content within limits when tested in accordance with ASTM D2709; and d. Verify that the new fuel oil has an absolute specific gravity at 60/60°F of > 0.83 and < 0.89 when tested in accordance with ASTM D1298 or an API gravity at 60°F of > 270 and < 390 when tested in accordance with ASTM D287 (Ref. 7).Failure to meet any of the above limits, except for clear and bright, is cause for rejecting the fuel oil, but does not represent a failure to meet the LCO concern since the fuel oil is not added to the storage'tanks.
If the fuel oil fails on clear and bright, it may be accepted if it passes water and sediment.
The specifications for water and sediment recognize that a small amount of water and sediment is acceptable.
Thus, this test may be used after a clear and bright test to provide a more quantitative result.Within 31 days following the initial new fuel oil sample, the fuel oil is analyzed to establish that the other properties specified in Table 1 of ASTM D975 (Ref. 7) are met for new fuel oil when tested in accordance with ASTM D975 (Ref. 7). The 31 day period is acceptable because the fuel oil properties of interest, even if they were not within stated limits, would not have an immediate effect on DG operation.
This Surveillance ensures the availability of high quality fuel oil for the DGs.Fuel oil degradation during long term storage shows up as an increase in particulate, due mostly to oxidation.
The presence of particulate does not mean the fuel oil will not burn properly in a diesel engine. The particulate can cause fouling of filters and fuel oil injection equipment, however, which can cause engine failure.Particulate concentrations should be determined based on ASTM D6217 (Ref. 7). This test method is used for assessing the mass quantity of Catawba Units 1 and 2 B 3.8.3-6 Revision No. 2 Diesel Fuel Oil, Lube Oil and Starting Air B 3.8.3 F :,.[S Stll1,Vl:H t ANCE RILOIJREMENTS (continued) particulates in middle distillate fuels, which includes 2-D diesel fuel. This method involves a gravimetric determination of total particulate concentration in the fuel oil and has a limit of 10 mgfl. For those designs in which the total stored fuel oil volume is contained in two or more interconnected tanks, each tank must be considered and tested separately.
The Frequency of this test takes into consideration fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between Frequency intervals.
SR 3.8.3.4 This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each DG is available.
The system design requirements provide for a minimum of five engine start cycles without recharging.
A start cycle is defined by the DG vendor, but usually is measured in terms of time (seconds of cranking) or engine cranking speed. The pressure specified in this SR is intended to reflect the lowest value at which the five starts can be accomplished.
Thbe 31 day Fre c takes into capacity, cap,'Sredundanc dand diversity of the o_.,urces and othe' cations/"lavaila the control ro" ,tigalrsoa frt the ope ra4 to_/Lbe fW normal air start pr ~sur SR 3.8.3.5 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel storage tanks<5 v Idý .eliminates the necessary environment for bacterial survival.
This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation.
Water may come from any of several sources, including condensation, ground water, rain water, and contaminated fuel oil, and from breakdown of the fuel oil-.by bacteria.
Frequent checking for and removal of accumulated water.minimizes fouling and rovides data regarding the watertight integrity of the fuel oil system The n e Fr re est fled~1----
Catawba Units 1 and 2 B 3.8.3-7 Revision No Diesel Fuel Oil, Lube Oil and Starting Air B 3.8.3 BASES SURVEILLANCE REQUIREMENTS (continued)
The presence of water does not necessarily represent failure of this SR, provided the accumulated water~is removed during performance of the ,,,
._.REFERENCES
- 1. UFSAR, Section 9.5.4.2.2. Regulatory Guide 1.137.3. ANSI N195-1976, Appendix B.4. UFSAR, Chapter 6.5. UFSAR, Chapter 15.6. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 7. ASTM Standards:
D4057; D975; D1298; D4176; D2709; D6217;and D287.8. UFSAR, Section 18.2.4.9, Catawba License Renewal Commitments, CNS-1274.00-00-0016, Section 4.5.Catawba Units 1 and 2 B 3.8.3-8 Revision No.9)
DC Sources-Operating B 3.8.4 BASES ACTIONS (continued) the loss of the channel DC power and the associated DG DC power, the load center power for the train is inoperable and the Condition(s) and Required Action(s) for the Distribution Systems must be entered immediately.
SURVEILLANCE REQUIREMENTS SR 3.8.4.1 Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the 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 conistent with the initia aes assumed in the batter sizing calculations.
The 7 Fr; is co ent with man rer 7recom-rpef oations andLý ,t4SEAT Z SR 3.8.4.2 Not used.SR 3.8.4.3 For the DC channel and DG batteries, visual inspection to detect corrosion of the battery terminals and connections, or measurement of the resistance of each intercell, interrack, intertier, and terminal connection, provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance.
The presence of visible corrosion does not necessarily represent a failure of this SR, provided an evaluation determines that the visible corrosion does not affect the OPERABILITY of the battery.Catawba Units 1 and 2 B 3.8.4-5 Revision NC(O DC Sources-Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.4.4 For the DC channel and DG batteries, visual inspection of the battery cells, cell plates, and battery racks provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance.
The presence of physical damage or deterioration does not necessarily represent a failure of this SR, provided an evaluation determines that the physical damage or deterioration does not affect the OPERABILITY of the battery (its ability to perform its design function).
Operating experien as shown tha se components u y pass the-SR when pe ed at the 18 m Frequency.
The ore, the , Freque was concluded t e acceptable fromreliability sta oint.SR 3.8.4.5 and SR 3.8.4.6 IN, T 2-Visual inspection and resistance measurements of intercell, interrack, intertier, and terminal connections provide an indication of physical damage or abnormal deterioration that could indicate degraded battery condition.
The anticorrosion material, as recommended by the manufacturer for the batteries, is used to help ensure good electrical connections and to reduce terminal deterioration.
The visual inspection for corrosion is not intended to require removal of and inspection under each terminal connection.
The removal of visible corrosion is a preventive maintenance SR. The presence of visible corrosion does not necessarily represent a failure of this SR provided visible corrosion is removed during performance of SR 3.8.4.5.Catawba Units 1adsshow2B a3.8.- Romeiponents usuins tN SSR w~hen ormed at the 18 mon 4requency.
Therefpr-lhe ( ,ilcy w~as concluded to 4&1~cceptable fromm a reffabiiy Catawba Units 1 and 2 B 3.8.4-6 Revision No.fJ DC Sources-Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.4.7 This SR requires that each battery charger for the DC channel be capable of supplying at least 200 amps and at least 75 amps for the DG chargers.
All chargers shall be tested at a voltage of at least 125 V for_> 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. These requirements are based on the design capacity of the chargers (Ref. 4). According to Regulatory Guide 1.32 (Ref. 10), the battery charger supply is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences.
The minimum required amperes and duration ensures that these requirements can be satisfied.
The eillance Freque y is acceptable, given th nit conditions r ired to perfor test and the other ad istrative controls ead elate charger performan uring these 18 mog;4 interval n addition, this Fr uenc is intendedt be co istent with)exp ed fuel cycle lengths SR 3.8.4.8 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 vital battery's actual duty cycle is identified in calculation CNC-1381.05-00-0011,125 VDC Vital Instrumentation and Control Power System Battery and Battery Charger Sizing Calculation.
The test duty cycle is the actual duty cycle adjusted for the temperature correction factor for 60OF operation, and a design margin of typically 10 to 15% for load addition.
The minimum DC battery terminal voltage is determined through Calculation CNC-1 381.05-00-0149, 125 VDC Vital I&C Power System (EPL) Voltage Drop Analysis.
The DG battery's actual duty cycle is identified in calculation CNC-1381.05-00-0050, 125 VDC Diesel Generator Battery and Battery Charger Sizing Calculation.
The test duty cycle is the actual duty cycle adjusted for the temperature correction factor for 60OF operation, and a design margin of typically 10 to 15% for load addition.
The minimum DG battery terminal voltage is determined through Calculations CNC-1 381.05-00-0235, Unit 1 125 VDC Essential Diesel Power System (EPQ) Voltage Drop Analysis and CNC-1381.05-00-0236, Unit 2 125 VDC Essential Diesel Power System (EPQ)Voltage Drop Analysis. (Note: The duty cycle in the UFSAR is used for battery sizing and includes the temperature factor of 11%, a design margin of 15%, and an aging factor of 25%.)Catawba Units 1 and 2 B 3.8.4-7 Revision Noo DC Sources-Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued)
Exce r perf ing SR 3.8. .4for the DC annel batýes with the u on line, Surveillanc requencyrf18 mo consistent w e recomi endations of egulatoryGuide 1.32 which s tes that the attery servi test shoudbe performed uuin refuelions or at s e other o e wifnte rIs between ts, not to This SR is modified by two Notes. Note 1 allows the performance of a modified performance discharge test in lieu of a service test.The modified performance discharge test is a performance discharge test that is augmented to include the high-rate, short duration discharge loads (during the first minute and 11-to-12 minute discharge periods) of the service test. The duty cycle of the modified performance test must fully envelope the duty cycle of the service test if the modified performance discharge test is to be used in lieu of the service test. Since the ampere-hours removed by the high-rate, short duration discharge periods of the service test represents a very small portion of the battery capacity, the-test rate can be changed to that for the modified performance discharge test without compromising the results of the performance discharge test.The battery terminal voltage for the modified performance discharge test should remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rates of the duty cycle). This will often confirm the battery's ability to meet the critical periods of the load duty cycle, in addition to determining its percentage of rated capacity.
Initial conditions for the modified performance discharge test should be identical to those specified for a service test. The reason for Note 2 is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems.Catawba Units 1 and 2 B 3.8.4-8 Revision No.
DC Sources-Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.4.9 A battery performance discharge 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 determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.A battery modified performance discharge test is described in the Bases for SR 3.8.4.8. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.9; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.9 while satisfying the requirements of SR 3.8.4.8 at the same time.The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 9). This reference recommends that the battery be replaced if its capacity is below 80% of the manufacturer's 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.
h S 6e meconth If the battery shows degradation, or if the battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the A,5E~ef( Surveillance Frequency is reduced to 18 months. However (for DC vital batteries only), if the battery shows no degradation but has reached 85%of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity _> 100% of the manufacturer's rating. Degradation is indicated, according to IEEE-450 (Ref. 9), when the battery capacity drops by more than 10% relative to its average capacity on the previous performance tests or when it is > 10% below the manufacturer's ra t in e n w rec natin ,-l This SR is modriied by a Note which is applicable to the DG batteries only. The reason for the Note is that performing the Surveillance would perturb the associated electrical distribution system and challenge safety systems.Catawba Units 1 and 2 B 3.8.4-9 Revision No.
Battery Cell Parameters B 3.8.6 BASES ACTIONS (continued)
B.1 and B.2 With one or more batteries (DC batteries, DG batteries, or both) with one or more battery cell parameters outside the Category C limit for any connected cell, sufficient capacity to supply the maximum expected load requirement is not assured and the corresponding DC electrical power subsystem must be declared inoperable.
Additionally, other potentially extreme conditions, such as not completing the Required Actions of Condition A within the required Completion Time or average electrolyte temperature of representative cells falling below 60'F, are also cause for immediately declaring the associated DC electrical power subsystem inoperable per Required Action B.1.In addition, Required Action B.2 mandates that the appropriate LCO(s)must then be entered for the DG supported by the inoperable DC subsystem.
If the plant is in MODES 1 through 4, LCO 3.8.1, "AC Sources -Operating" is required to be entered. If the DG is required to support equipment during MODES 5 or 6 or movement of irradiated fuel assemblies, regardless of operating mode, LCO 3.8.2, "AC Sources -Shutdown," is the appropriate LCO.Required Action B.2 is modified by a Note indicating that it is only applicable for inoperable DG batteries.
SURVEILLANCE REQUIREMENTS SR 3.8.6.1 SR 3.8.6.2 Not used.Catawba Units 1 and 2 B 3.8.6-3 Revision N, .
Battery Cell Parameters B 3.8.6 BASES SURVEILLANCE REQUIREMENTS (continued)
.- .2-S R 3.8.6.3 The qua insp on of channels lC and DG tteries forgravi nd vo consistent with IEEE-4K1Jl n adi-o, wi in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a battery discharge
< 110 or a attery overcharge
> 150 V, the battery must be demonstrated to meet Category B limits. Transients, such as motor starting transients, which may momentarily cause battery voltage to drop to _< 110 V, do not constitute a battery discharge provided the battery terminal voltage and float current return to pre-transient values. This inspection is also consistent with IEEE-450 (Ref. 4), which recommends special inspections following a severe discharge or overcharge, to ensure that no significant degradation of the battery occurs as a consequence of such discharge or overcharge.
SR 3.8.6.4 This S eillance verificfai that the av age temperature re sentajv.c 1lsj_ <-60oF, is con
- tent with a reco endation of EE-450 , that states th he temperature o lectrolytes in representa,.
1 cells should b etermined on a q rterly basis.Lower than normal temperatures act to inhibit or reduce battery capacity.This SR ensures that the operating temperatures remain within an acceptable operating range. This limit is based on manufacturer recommendations.
The term "representative cells" replaces the fixed number of "six connected cells", consistent with the recommendations of IEEE-450 (Ref.4) to provide a general guidance to the number of cells adequate to monitor the temperature of the battery cells as an indicator of satisfactory performance.
For some cases, the number of cells may be less than six, in other conditions, the number may be more.Catawba Units 1 and 2 B 3.8.6-4 Revision N I nverters--Operating B 3.8.7 BASES ACTIONS (continued)
Required Action A.1 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to fix the inoperable inverter and return it to service. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limit is based upon engineering judgment, taking into consideration the time required to repair an inverter and the additional risk to which the unit is exposed because of the inverter inoperability.
This has to be balanced against the risk of an immediate shutdown, along with the potential challenges to safety systems such a shutdown might entail. When the AC vital bus is powered from its voltage regulated transformer, it is relying upon interruptible AC electrical power sources (offsite and onsite). The uninterruptible inverter source to the AC vital buses is the preferred source for powering instrumentation trip setpoint devices.If the channel-related inoperable inverter is replaced by its train's swing inverter, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limit does not apply (unless the swing inverter is also inoperable).
B.1 and B.2 If the inoperable devices or components cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit 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 unit conditions from full power conditions in an orderly manner and without challenging plant systems.SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and AC vital bus energized from the inverter.
The verification of proper indicated voltage output ensures that the required power is readily available for the instrumentation of the RPS and ESFAS connecte2 B o the AC vital buses.R lhe 7 equvesn Noakes.o unt the3- ndant capabli t.e te inves nd; o~t:h7/I i~dcations j~able in th otroom tha. aert the op to inverterJ;__nafun/
...c ns , ." -Catawba Units 1 and 2 B 3.8.7-3 Revision I nverters-Shutdown B 3.8.8 BASES ACTIONS (continued) this option may involve undesired administrative efforts. Therefore, the allowance for sufficiently conservative actions is made (i.e., to suspend CORE ALTERATIONS, movement of irradiated fuel assemblies, and operations involving positive reactivity additions) that could result in loss of required SDM (MODE 5) or required boron concentration (MODE 6).Suspending positive reactivity additions that could result in failure to meet the minimum SDM or boron concentration limits is required to assure continued safe operation..
Introduction of coolant inventory must be from sources that have a boron concentration greater than that what would be required in the RCS for minimum SDM or refueling boron concentration.
This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.
Introduction of temperature changes including temperature increases when operating with a positive MTC must also be evaluated to ensure they do not result in a loss of required SDM.Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition.
These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required inverters and to continue this action until restoration is accomplished in order to provide the necessary inverter power to the unit safety systems.The Completion Time of immediately is consistent with the required times for actions requiring prompt attention.
The restoration of the required inverters should be completed as quickly as possible in order to minimize the time the unit safety systems may be without power or powered from a constant voltage source transformer.
SURVEILLANCE SR 3.8.8.1 REQUIREMENTS This Surveillance verifies that the power sources are functioning properly with all required circuit breakers closed and AC vital bus energized from the required power source. The verification of proper indicated voltage ensures that required power is readily available for the instrumentation connected to the AC vital bus. i-IIe 7xeaN, Fret6encv takes iMo account)Catawba Units 1 and 2 B 3.8.8-3 Revision Distribution Systems-Operating B 3.8.9 BASES ACTIONS (continued) status,, the unit 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 unit conditions from full power conditions in an orderly manner and without challenging plant systems.F..1 Condition F corresponds to a level of degradation in the electrical power distribution system that causes a required safety function to be lost.When more than one inoperable electrical power distribution subsystem results in the loss of a required function, the plant is in a condition outside the accident analysis.
Therefore, no additional time is justified for continued operation.
LCO 3.0.3 must be entered immediately to commence a controlled shutdown.SURVEILLANCE SR 3.8.9.1 REQUIREMENTS This Surveillance verifies that the AC, channels of DC, DC trains, and AC vital bus electrical power distribution systems are functioning properly, with the correct circuit breaker alignment.
The correct breaker alignment ensures the appropriate separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required bus. The verification of proper indicated voltage availability on the buses ensures that the required voltage is readily available for motive as well as control functions for critical system loads connected to these bue.Te 7 daFrequency takes int pccodunffte-o/-f the AC,D, and AC vital bus e rical power distribýW ubsys;e s, and ot indications availa tirQmthatfalert theoperator/
to sytem REFERENCES
- 1. UFSAR, Chapter 6.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 4. Regulatory Guide-1.93, December 1974.Catawba Units 1 and 2 B 3.8.9-9 Revision Distribution Systems-Shutdown B 3.8.10 BASES SURVEILLANCE SR 3.8.10.1 REQUIREMENTS This Surveillance verifies that the AC, channels of DC, DC trains, and AC vital bus electrical power distribution subsystems are functioning properly, with all the buses energized.
The verification of proper indicated 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. 7 day Frequen taes into a.p.oeount the-ýiaiýý fte ectnca ;zwer distributio 5Obsystems, other v (indicatio available in e. ontrol roor vtat aler the Zlerator to/subs ~em malfunc *"ns. i-REFERENCES
- 1. UFSAR, Chapter 6.2. UFSAR, Chapter 15.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.8.10-4 Revision NO CHANGES THIS PACE.O INFORMATiOw ONLY Boron Concentration B 3.9.1 BASES ACTIONS A.1 and A.2 Continuation of CORE ALTERATIONS or positive reactivity additions (including actions to reduce boron concentration) is contingent upon maintaining the unit in compliance with the LCO. If the boron concentration of any coolant volume in the RCS, the refueling canal, or the refueling cavity is less than its limit, all operations involving CORE ALTERATIONS or positive reactivity additions must be suspended immediately.
Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position.
Operations that individually add limited positive reactivity (e.g., temperature fluctuations from inventory addition or temperature control fluctuations), but when combined with all other operations affecting core reactivity (e.g., intentional boration) result in overall net negative reactivity addition, are not precluded by this Action.A.3 In addition to immediately suspending CORE ALTERATIONS and positive reactivity additions, boration to restore the concentration must be initiated immediately.
In determining the required combination of boration flow rate and concentration, no unique Design Basis Event 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 unit conditions.
An acceptable method is to borate at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent.
Once actions have been initiated, they must be continued until the boron concentration is restored.
The restoration time depends on the amount of boron that must be injected to reach the required concentration.
SURVEILLANCE SR 3.9.1.1 REQUIREMENTS This SR ensures that the coolant boron concentration in the RCS, and connected portions of the refueling canal and the refueling cavity, is within the COLR limits. The boron concentration of the coolant in each required v.lume is determined periodically by chemical analysis.
Priorto re-connecting portions of the refueling canal or the refueling cavity to the RCS, this SR must be met per SR 3.0.4. If any dilution activity has Catawba Units 1 and 2 B 3.9.1-3 Revision No. 2 Boron Concentration B3.9.1 BASES SURVEILLANCE REQUIREMENTS (continued) occurred while the cavity or canal were disconnected from the RCS, this SR ensures the correct boron concentration prior to communication with the RCS. One sample from the refueling canal or reactor cavity is sufficient to determine the boron concentration in that volume of water.An additional sample is taken from the RCS.Fr quency of onc very 72;A the boron copýentration of j REFERENCES
- 1. 10 CFR 50, Appendix A, GDC 26.2. 10 CFR 50.36, Technical Specifications (c)(2)(ii).
Catawba Units 1 and 2 B 3.9.1-4 Revision No.V Nuclear Instrumentation B 3.9.2 BASES SURVEILLANCE REQUIREMENTS SR 3.9.2.1 SR 3.9.2.1 is the performance of a CHANNEL CHECK which is a comparison of the parameter indicated on one channel to a similar parameter on other channels.
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 channels, but each channel should be consistent with its local conditions.
[The Freqdn'cy of 12 houi SR 3.9.2.2 SR is the performance of a CHANNEL CALIBRATION2EZ Aa This SR is modified by a Note stating that neutron detector sensors (NIS and BDMS) are excluded from the CHANNEL CALIBRATION.
The CHANNEL CALIBRATION for the source range neutron flux monitors (NIS) consists of obtaining the detector plateau and pulse height discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data.The CHANNEL CALIBRATION for the source range neutron flux monitors (Gamma-Metrics) consists of verifying that the channels respond correctly to test inputs with the necessary range and accuracy.h-718 month quen ixs d on the need t erform t.4'urveillanc rnder the con ons that apply ng, a pla outage.Operati experience ha hown these co ponents all ass the S ance when pe rmed at the 1 .onth Fre_ -ncy REFERENCES
- 1. 10 CFR 50, Appendix A, GDC 13, GDC 26, GDC 28, and GDC 29.2. UFSAR, Sections 4.2, 15.4.6.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.9.2-3 Revision NO-0 Containment Penetrations B 3.9.3 BASES APPLICABILITY (continued) not completely block the penetration or be capable of resisting pressure.The purpose is to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper directions such that it can be treated and monitored.
ACTIONS A.1 and A.2 If the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, the unit must be placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending movement of recently irradiated fuel assemblies within containment.
Performance of these actions shall not preclude completion of movement of a component to a safe position.B.1 and B.2 With one or more Containment Purge Exhaust System heaters inoperable, the heater must be restored to OPERABLE status within 7 days. Alternatively, a report must be initiated per Specification 5.6.6, which details the reason for the heater's inoperability and the corrective action required to return the heater to OPERABLE status.The heaters do not affect OPERABILITY of the Containment Purge Exhaust System filter trains because carbon adsorber efficiency testing is performed at 306C and 95% relative humidity.
The accident analysis shows that site boundary radiation doses are within the limits of 10 CFR 50.67 and Regulatory Guide 1.183 during a DBA LOCA under these conditions.
SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position.
The Surveillance on the open purge and exhaust valves will demonstrate that the valves are exhausting through an OPERABLE Containment Purge Exhaust System HEPA Filter and carbon adsorber.he Su.. nce is performed ev ays during movement of ntly irrad fuel assembliwn containment.
The Surva 2e interval is/ ,811ected to be co surate with the normal duratio .rtime to complete/ uel handlinog.
erations.
As such, this Surveilla oe. ensures that a "_ postulate tfuel handling accident involving recently irradiated fuel thatJ Catawba Units 1 and 2 B 3.9.3-4 Revision Containment Penetrations B 3.9.3 BASES SURVEILLANCE REQUIREMENTS (continued) release$ssion product dioactivity.withj thecont ent will t result inea rease of signific t fission prod radioaOyit to the e ironme SR 3.9.3.2 Standby systems should be checked periodically to ensure that they function properly s nvi nment an nto _.r.g conditi0ns on}no] sev eesting each trai e month-provL~ies ansa ThiseSRaverifiecs th0t is system. M eatperforatimei ry out any wt SoiVre that m Fy have a ccumulaed in the carbon from humidity in the Exhaus .Systems with heaters n ardust ne operated by initifatinc through the HEPA filters and activated carbon adsorbers for > 1 continuous hours with the heaters of. tay Fre Yasc carso hene n re anflin ispmenitf n e two traedundanci t es SR 3.9.3.3 This SR verifies that the required testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The Containment Purge Exhaust System filter tests are in accordance with Reference
- 4. The VFTP includes testing HEPA filter performance, carbon adsorbers efficiency, system flow rate, and the physical properties of the activated carbon (general use and following specific operations).
Specific test Frequencies and additional information are discussed in detail in the VFTP.REFERENCES
- 1. UFSAR, Section 15.7.4.2. NUREG-0800, Section 15.7.4, Rev. 1, July 1981.3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
- 4. Regulatory Guide 1.52 (Rev. 2).5. 10 CFR 50.67, Accident source term.6. Regulatory Guide 1.183 (Rev. 0).7. Catawba Nuclear Station License Amendments 90/84 for Units 1/2, August 23, 1991.Catawba Units 1 and 2 B 3.9.3-5 Revision Noj' RHR and Coolant Circulation-High Water Level B 3.9.4 BASES ACTIONS (continued)
A.3 If RHR loop requirements are not met, actions shall be initiated and continued in order to satisfy RHR loop requirements.
With the unit in MODE 6 and the refueling water level _ 23 ft above the top of the reactor vessel flange, corrective actions shall be initiated immediately.
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 Surveillance demonstrates that the RHR loop is in operation and circulating, reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. The RCS temperature is detrmined to an is aitaied.
of oTHurs is sufficiepýcnsie.o e-fow, tempe t_ rb, pump con~l.e and alarmn indi ons availa he....e controlro.
for monitoring t telHR -S Sstem- _ " --REFERENCES
- 1. UFSAR, Section 5.5.7.2. 10 CFR 50.36, Technical Specifications, (cX2)(ii).
- 3. NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management." Catawba Units 1 and 2 B 3.9.4-4 Revision No RHR and Coolant Circulation-Low Water Level B 3.9.5 BASES ACTIONS (continued) concentration greater than that which would be required in the RCS for minimum refueling boron concentration.
This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.
B.2 If no RHR loop is in operation, actions shallbe initiated immediately, and continued, to restore one RHR loop to operation.
Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously.
B.3 If no RHR loop is in operation, 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 that 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 appropriate for the majority of time during refueling operations, based on time to coolant boiling, since water level is not routinely maintained at low levels.SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability, prevent vortexing in the suction of the RHR pumps, and to prevent thermal and boron stratification in the core. The RCS temperature is determined to ensure the appropriate decay heat removal is maintained.
In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR umo suction requirementsmust be metu aTtae wraeque y of 12hours iUsufficsent, cor1nderin the 95w, temvs pu p conit ol, and alarm i ava/lab~le to t ....lnitori' g the RHR SYSt, m in the control roorn Catawba Units 1 and 2 B 3.9.5-3 Revision RHR and Coolant Circulation-Low Water Level B 3.9.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.9.5.2 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 veri{ proper breaker aligrment and power available the reouired Dum he Frequenqorf 7 days is consideed reason in l of er cprol availab_ amd'fasbeen_
n to be REFERENCES
- 1. UFSAR, Section 5.5.7.2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).
Catawba Units 1 and 2 B 3.9.5-4 111q_Revision No. 1, Refueling Cavity Water Level B 3.9.6 BASES APPLICABILITY LCO 3.9.6 is applicable during CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, and is also applicable when moving irradiated fuel assemblies within containment.
The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis.
If irradiated fuel assemblies are not present in containment, there can be no significant radioactivity release as a result of a postulated fuel handling accident.Requirements for fuel handling accidents in the spent fuel -pool are covered by LCO 3.7.14, "Fuel Storage Pool Water Level." ACTIONS A.1 and A.2 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 REQUIREMENTS SR 3.9.6.1 Verification of a minimum water level of 23 ft above the top of the react.vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met.Water 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. 2).The Fre ency of 24 h/urs is based on ýeineering judgme nd is ,uons ered adequat in view of the lar volume of water nd the normal 2 prcedural controgof valve position.
-which make si ificant unplannenj',level changes),nlikely.J-
._Catawba Units 1 and 2 B 3.9.6-2 Revision Nov)
Unborated Water Source Isolation Valves B 3.9.7 BASES SURVEILLANCE REQUIREMENTS SR 3.9.7.1 These valves are to be secured closed to isolate possible dilution paths.The likelihood of a significant reduction in the boron concentration during MODE 6 operations is remote due to the large mass of borated water in the refueling cavity and the fact that all unborated water sources are is d precluding a dilution.
The boron concentration is drs~during MODE 6 under SR3.9.1.1.
This Surveillance d-mons-thiates that the valves are closed through a system walkdown.lhe 1 a requen yis based on gineeri udgment~rfd is consi d re onable iL i-ew of othe ministr ls4at will e rre th-. e alve opo ein'q is an un evly possibility.
- REFERENCES
- 1. UFSAR, Section 15.4.6.2. NUREG-0800, Section 15.4.6.Catawba Units 1 and 2 B 3.9.7-3 Revis I i on N .00 ATTACHMENT 5 TSTF-425, Rev.3 (NUREG-1431)
SURVEILLANCES VERSUS CATAWBA SURVEILLANCES CROSS REFERENCE TABLE ATTACHMENT 5 TSTF-425, Rev. 3 (NUREG-1431)
SRs vs. Catawba SRs Cross Reference TSTF SR Catawba COMMENTS SR&, Bases 1.1 1.1 Deleted in TS and relocated to the Surveillance Frequency Control Program (SFCP).3.1.1.1 3.1.1.1 3.1.2.1 3.1.2.1 3.1.4.1 3.1.4.1 3.1.4.2 3.1.4.2 3.1.5.1 3.1.5.1 3.1.6.2 3.1.6.2 3.1.6.3 3.1.6.3 3.1.8.2 3.1.8.2 3.1.8.3 3.1.8.3 3.1.8.4 3.1.8.4 3.2.1.1 3.2.1.1 Catawba used a modified Option B. (RAOC)3.2.1.2 3.2.1.2 Catawba used a modified Option B. (RAOC)3.2.1.2 3.2.1.3 Catawba used a modified Option B (RAOC). (This is an additional SR that encompasses TSTF SR 3.2.1.2.)3.2.2.1 3.2.2.1 3.2.2.2 Catawba SR not in TSTF.3.2.3.1 3.2.3.1 Catawba used a modified Option B. (RAOC)3.2.4.1 3.2.4.1 3.2.4.2 3.2.4.2 3.3.1.1 3.3.1.1 3.3.1.2 3.3.1.2 3.3.1.3 3.3.1.3 3.3.1.4 3.3.1.4 A note concerning an expired one-time extension on a surveillance is deleted here, also.3.3.1.5 3.3.1.5 3.3.1.6 3.3.1.6 3.3.1.7 3.3.1.7 7/1/09 LAR modifies the Bases only for this SR. The Catawba SR will be relocated here, also.3.3.1.8 3.3.1.8 7/1/09 LAR modifies the Bases only for this SR. The Catawba SR will be relocated here, also.3.3.1.9 3.3.1.9 3.3.1.10 3.3.1.10 3.3.1.11 3.3.1.11 7/1/09 LAR proposes to modify the SR description.
SR 3.3.1.11 will be relocated here, also.3.3.1.12 3.3.1.12 3.3.1.13 3.3.1.13 3.3.1.14 3.3.1.14 3.3.1.16 3.3.1.16 3.3.1.17 Catawba SR not in TSTF.3.3.2.1 3.3.2.1 Page 1 of 11 TSTF SR Catawba COMMENTS SR&-Bases 3.3.2.2 3.3.2.2 3.3.2.3 TSTF SR not in Catawba TS.3.3.2.3 Catawba SR not in TSTF.3.3.2.4 3.3.2.4 3.3.2.5 3.3.2.5 3.3.2.6 3.3.2.6 3.3.2.7 -------- TSTF SR not in Catawba TS.3.3.2.7 Catawba SR not in TSTF. Note: 9/2/08 LAR modifies the Bases for this SR.3.3.2.8 3.3.2.8 3.3.2.9 3.3.2.9 9/2/08 LAR modifies the Bases for this SR.3.3.2.10 3.3.2.10 3.3.2.11 TSTF SR not in Catawba TS.3.312.11 Catawba SR not in TSTF.3.3.2.12 Catawba SR not in TSTF.3.3.3.1 3.3.3.1 3.3.3.32 Catawba SR 3.3.3.2 is an unused SR.3.3.3.2 3.3.3.3 3.3.4.1 3.3.4.1 3.3.4.2 ---------
TSTF SR not in Catawba TS.3.3.4.3 3.3.4.2 3.3.4.4 -------- TSTF SR not in Catawba TS.3.3.5.1 TSTF SR not in Catawba TS.3.3.5.2 3.3.5.1 3.3.5.3 3.3.5.2 3.3.6.1 TSTF SR not in Catawba TS.3.3.6.2 TSTF SR not in Catawba TS.3.3.6.3 TSTF SR not in Catawba TS.3.3.6.4 3.3.6.1 3.3.6.5 3.3.6.2 3.3.6.6 TSTF SR not in Catawba TS.3.3.6.7 3.3.6.3 A 9/2/08 LAR modifies the Bases only for this SR. The SR will be relocated here, also.3.3.6.8 3.3.6.4 A 9/2/08 LAR modifies the TS SR and the Bases. The SR will be relocated here, also.3.3.6.9 ----------
TSTF SR not in Catawba TS.3.3.7.1 ------- Catawba does not have this TSTF TS.3.3.7.2 Catawba does not have this TSTF TS.3.3.8.1 Catawba does not have this TSTF TS.3.3.8.2 Catawba does not have this TSTF TS.3.3.8.3 Catawba does not have this TSTF TS.3.3.9.1 3.3.9.1 3.3.9.2 Catawba SR not in TSTF.,3.3.9.3 Catawba SR not in TSTF.3.3.9.4 Catawba SR not in TSTF.3.3.9.5 Catawba SR not in TSTF.I Page 2 of 11 TSTF SR Catawba COMMENTS SR&Bases 3.3.9.2 3.3.9.6 3.3.9.3 TSTF SR not in Catawba TS.3.4.1.1 3.4.1.1 3.4.1.2 3.4.1.2 3.4.1.3 3.4.1.3 3.4.1.4 TSTF SR not in Catawba TS.3.4.1.4 Catawba SR not in TSTF.3.4.2.1 ----------
Event driven SR at Catawba, retain frequency in TS.3.4.3.1 3.4.3.1 3.4.4.1 3.4.4.1 3.4.5.1 3.4.5.1 3.4.5.2 3.4.5.2 3.4.5.3 3.4.5.3 3.4.6.1 3.4.6.1 3.4.6.2 3.4.6.2 3.4.6.3 3.4.6.3 3.4.7.1 3.4.7.1 3.4.7.2 3.4.7.2 3.4.7.3 3.4.7.3 3.4.8.1 3.4.8.1 3.4.8.2 3.4.8.2 3.4.9.1 3.4.9.1 3.4.9.2 3.4.9.2 3.4.9.3 3.4.9.3 3.4.11.1 3.4.11.1 Page 3 of 11 TSTF SR Catawba COMMENTS SR&Bases 3.4.11.2 3.4.11.2 3.4.11.3 Catawba SR not in TSTF.3.4.11.3----------
TSTF SR not in Catawba TS.3.4.11.4 ----------
TSTF SR not in Catawba TS.3.4.12.1 3.4.12.1 Catawba SR is TSTF SR 3.4.12.1 and 3.4.12.2 combined.3.4.12.2 ---------
See above.3.4.12.3 3.4.12.2 3.4.12.4 3.4.12.3 3.4.12.5 ----------
TSTF SR not in Catawba TS.3.4.12.6 3.4.12.4 3.4.12.7 3.4.12.7 3.4.12.8 3.4.12.5 3.4.12.9 3.4.12.6 3.4.13.1 3.4.13.1 3.4.13.2 3.4.13.2 3.4.14.1 3.4.14.1 3.4.14.2 3.4.14.2 3.4.14.3 ----------
TSTF SR not in Catawba TS.3.4.15.1 3.4.15.1 3.4.15.2 3.4.15.2 3.4.15.3 3.4.15.3 3.4.15.4 3.4.15.4 3.4.15.5 3.4.15.5 3.4.15.6 Catawba SR not in TSTF.Page 4 of 11 TSTF SR Catawba COMMENTS SR&Bases 3.4.16.1 3.4.16.1 A 12/15/09 LAR proposes to modify the SR description to allow Xe-133 sampling.
This SR will be relocated here, also.3.4.16.2 3.4.16.2 3.4.16.3 3.4.16.3 A 12/15/09 LAR proposes to modify Catawba SR 3.4.16.3.
This SR will be relocated here, also.3.4.17.2 TSTF SR not in Catawba TS. [RCS Loops Isolation Valves].3.4.19.1 3.4.17.1 3.4.19.2 3.4.17.2 TSTF does not revise this SR. Catawba is not relocating this SR.3.4.19.3 TSTF does not revise this SR.3.5.1.1 3.5.1.1 3.5.1.2 3.5.1.2 3.5.1.3 3.5.1.3 3.5.1.4 3.5.1.4 3.5.1.5 3.5.1.5 3.5.2.1 3.5.2.1 3.5.2.2 3.5.2.2 3.5.2.3 3.5.2.3 3.5.2.5 3.5.2.5 3.5.2.6 3.5.2.6 3.5.2.7 3.5.2.7 3.5.2.8 3.5.2.8 3.5.4.1 3.5.4.1 3.5.4.2 3.5.4.2 9/2/08 LAR proposes to modify the minimum RWST volume. SR 3.5.4.2 will be relocated here, also.3.5.4.3 3.5.4.3 3.5.5.1 3.5.5.1 3.5.6 TSTF SR not in Catawba TS.3.6.2.1 3.6.2.1 TSTF does not revise this SR. Catawba is not relocating this SR.3.6.2.2 3.6.2.3 3.6.2.2 Catawba SR not in TSTF.3.6.3.1 3.6.3.1 3.6.3.2 3.6.3.2 3.6.3.3 3.6.3.3 Page 5 of 11 TSTF SR Catawba COMMENTS SR&Bases 3.6.3.4 3.6.3.4 TSTF does not revise this SR. Catawba is not relocating this SR.3.6.3.5 3.6.3.5 The SR is only based on the In-service Testing Program. Catawba SR 3.6.3.5 is not revised.3.6.3.6 TSTF SR not in Catawba TS.3.6.3.6 Catawba SR 3.6.3.6 will not be relocated since it is based on the Containment Leakage Rate Testing Program.3.6.3.8 3.6.3.7 3.6.3.9 ------ TSTF SR not in Catawba TS.3.6.3.10 TSTF SR not in Catawba TS.3.6.3.11 3.6.3.8 TSTF does not revise this SR. Catawba is not relocating this SR.3.6.4.1 A 3.6.4.1 3.6.4B from the TSTF is not applicable.
3.6.5.1 B 3.6.5.1 3.6.5A and 3.6.5C from the TSTF are not applicable.
3.6.5.2 B 3.6.5.2 3.6.6.1 C 3.6.6.1 3.6.6A, B, D and E from the TSTF are not applicable.
3.6.6.2 C 3.6.6.2 TSTF does not revise this SR. Catawba is not relocating this SR.3.6.6.3 C 3.6.6.3 3.6.6.4 C 3.6.6.4 9/2/08 LAR proposes to modify this SR description.
SR 3.6.6.4 will be relocated here, also.-- 3.6.6.5 Catawba SR not in TSTF.3.6.6.6 Catawba SR not in TSTF.3.6.6.5.C 3.6.6.7 A 9/30/09 License Amendment proposes to modify the frequency of this SR.3.6.7 ----------
Catawba does not have this TSTF TS.3.6.8.1 ----------
TSTF SR not in Catawba TS.Page 6 of 11 TSTF SR Catawba SR&Bases COMMENTS 3.6.8.2 3.6.16.1 3.6.16.2 Catawba SR not in TSTF.3.6.8.3 3.6.16.3 TSTF does not modify the SR. However, it is relocated here since Catawba has a frequency of "3 times in 10 years" specified.
3.6.8.4 -----------
TSTF SR not in Catawba TS.3.6.9.1+ 3.6.8.1 3.6.9.2 TSTF SR not in Catawba TS.3.6.8.2 Catawba SR not in TSTF.3.6.8.3 Catawba SR not in TSTF.3.6.9.3 3.6.8.4 3.6.10.1 3.6.9.1 3.6.10.2 3.6.9.2 3.6.10.3 3.6.9.3 3.6.11 ------ TSTF TS not applicable at Catawba.3.6.13.1 3.6.10.1 3.6.13.2 3.6.10.2 TSTF does not revise this SR. Catawba is not relocating this SR.3.6.13.3 3.6.10.3 3.6.13.4 3.6.10.4 3.6.13.5 3.6.10.5 3.6.10.6 Catawba SR not in TSTF.3.6.14.1 3.6.11.1 3.6.14.2 3.6.11.2 3.6.14.3 TSTF SR not in Catawba TS.3.6.14.4 3.6.11.3 3.6.11.4 Catawba SR not in TSTF.Page 7 of 11 TSTF SR Catawba COMMENTS SR &Bases 3.6.11.5 Catawba SR not in TSTF.3.6.11.6 Catawba SR not in TSTF.3.6.11.7 Catawba SR not in TSTF.3.6.15.1 3.6.12.1 3.6.15.2 3.6.12.4 3.6.15.3 3.6.12.5 3.6.15.4 3.6.12.3 3.6.15.5 3.6.12.7 3.6.15.6 3.6.12.6 3.6.15.7 3.6.12.2 TSTF does not revise this SR. Catawba is not relocating this SR.3.6.16.1 3.6.13.1 10/2/08 LAR proposes to modify this SR. Catawba SR will be relocated here, also.3.6.16.2 3.6.13.2 3.6.16.3 3.6.13.4 10/2/08 LAR proposes to modify this SR. Catawba SR will be relocated here, also.3.6.16.4 3.6.13.5 10/2/08 LAR proposes to modify this SR. Catawba SR will be relocated here, also.3.6.16.5 3.6.13.6 10/2/08 LAR proposes to modify Catawba SR 3.6.13.6.
This SR will be relocated here, also.3.6.16.6 3.6.13.7 3.6.16.7 3.6.13.3 3.6.17.1 3.6.14.1 TSTF does not modify this SR. Catawba is not relocating this SR.3.6.17.2 3.6.14.2 3.6.17.3 3.6.14.3 TSTF does not modify this SR. Catawba is not relocating this SR.3.6.17.4 3.6.14.4 3.6.17.5 3.6.14.5 3.6.15.1 Event-driven SR and thus, not relocated.
3.6.18.1 3.6.15.2 Catawba SR is TSTF SR 3.6.18.1 c. only.Page 8 of 11 TSTF SR Catawba COMMENTS SR &Bases 3.6.18.2 3.6.15.3 3.7.1.1 TSTF does not modify this SR. Catawba is not relocating this SR.3.7.2.2 ----------
TSTF SR not in Catawba TS. (SR 3.7.2.1 is not revised by the TSTF)3.7.3.2 ------- TSTF SR not in Catawba TS. (SR 3.7.3.1 is not revised by the TSTF)3.7.4.1 Catawba SR not in TSTF.3.7.4.1 3.7.4.2 3.7.4.2 3.7.4.3 3.7.5.1 3.7.5.1 3.7.5.2 3.7.5.2 TSTF does not modify this SR. Catawba is not relocating this SR.3.7.5.3 3.7.5.3 3.7.5.4 3.7.5.4 3.7.5.5 3.7.5.5 TSTF does not modify this SR. Catawba is not relocating this SR.3.7.6.1 3.7.6.1 3.7.7.1 3.7.7.1 3.7.7.2 3.7.7.2 3.7.7.3 3.7.7.3 3.7.8.1 3.7.8.1 3.7.8.2 3.7.8.2 3.7.8.3 3.7.8.3 3.7.9.1 3.7.9.1 3.7.9.2 3.7.9.2 3.7.9.3 Catawba SR not in TSTF.3.7.9.3 TSTF SR not in Catawba TS.3.7.9.4 TSTF SR not in Catawba TS.3.7.10.1 3.7.10.1 3.7.10.2 3.7.10.2 TSTF does not modify this SR. Catawba is not relocating this SR.3.7.10.3 3.7.10.3 3.7.10.4 3.7.10.4 Catawba SR Frequency is specified by the CRH Program in TS 5.5.16.Catawba SR was not relocated.
3.7.11.1 ----------
TSTF SR not in Catawba TS.3.7.11.1 Catawba SR not in TSTF.3.7.12.1 3.7.12.1 3.7.12.2 3.7.12.2 TSTF does not modify this SR. Catawba is not relocating this SR.3.7.12.3 3.7.12.3 3.7.12.4 3.7.12.4 3.7.12.5 TSTF SR not in Catawba TS.3.7.13.1 Catawba SR not in TSTF.3.7.13.1 3.7.13.2 3.7.13.2 3.7.13.3 TSTF does not modify this SR.3.7.13.3 TSTF SR not in Catawba TS.3.7.13.4 3.7.13.4 3.7.13.5 3.7.13.5 3.7.14 Catawba does not have this TSTF TS.3.7.15.1 3.7.14.1 Page 9 of 11 TSTF SR Catawba COMMENTS SR &Bases 3.7.16.1 3.7.15.1 3.7.18.1 3.7.17.1 3.8.1.1 3.8.1.1 3.8.1.2 3.8.1.2 5/28/09 LAR proposes to modify the SR Description to change the steady state voltage values. The Catawba SR will be relocated here, also.3.8.1.3 3.8.1.3 3.8.1.4 3.8.1.4 3.8.1.5 3.8.1.5 3.8.1.6 3.8.1.6 3.8.1.7 3.8.1.7 5/28/09 LAR proposes to modify the SR Description to change the steady state voltage values. The Catawba SR will be relocated here, also.3.8.1.8 3.8.1.8 3.8.1.9 3.8.1.9 5/28/09 LAR proposes to modify the SR Description to change the steady state voltage values. The Catawba SR will be relocated here, also.3.8.1.10 3.8.1.10 3.8.1.11 3.8.1.11 5/28/09 LAR proposes to modify the SR Description to change the steady state voltage values. The Catawba SR will be relocated here, also.3.8.1.12 3.8.1.12 5/28/09 LAR proposes to modify the SR Description to change the steady state voltage values. The Catawba SR will be relocated here, also.3.8.1.13 3.8.1.13 3.8.1.14 3.8.1.14 3.8.1.15 3.8.1.15 5/28/09 LAR proposes to modify the SR Description to change the steady state voltage values. The Catawba SR will be relocated here, also.3.8.1.16 3.8.1.16 3.8.1.17 3.8.1.17 3.8.1.18 3.8.1.18 3.8.1.19 3.8.1.19 5/28/09 LAR proposes to modify the SR Description to change the steady state voltage values. The Catawba SR will be relocated here, also.3.8.1.20 3.8.1.20 5/28/09 LAR proposes to modify the SR Description to change the steady state voltage values. The Catawba SR will be relocated here, also.3.8.3.1 3.8.3.1 3.8.3.2 3.8.3.2 3.8.3.3 3.8.3.3 TSTF does not modify this SR. Catawba is not relocating this SR.3.8.3.4 3.8.3.4 3.8.3.5 3.8.3.5 3.8.4.1 3.8.4.1 3.8.4.2 Catawba SR is labeled "not used". Catawba SR not in TSTF.3.8.4.3 Catawba SR not in TSTF. A 12-14-09 LAR proposes to modify this SR to refer to a new Table 3.8.4-1. The Catawba SR will be relocated here, also.3.8.4.4 Catawba SR not in TSTF.Page 10 of 11 TSTF SR Catawba COMMENTS SR&Bases 3.8.4.5 Catawba SR not in TSTF.3.8.4.6 Catawba SR not in TSTF. A 12-14-09 LAR proposes to modify this SR to refer to a new Table 3.8.4-1. The Catawba SR will be relocated here, also.3.8.4.2 3.8.4.7 3.8.4.3 3.8.4.8 3.8.4.9 Catawba SR not in TSTF.3.8.6.1 -----------
TSTF SR not in Catawba TS.3.8.6.2 -----------
TSTF SR not in Catawba TS.3.8.6.3 -----------
TSTF SR not in Catawba TS.3.8.6.4 3.8.6.4 3.8.6.5 -----------
TSTF SR not in Catawba TS.3.8.6.6 -----------
TSTF SR not in Catawba TS.3.8.6.1 Catawba SR not in TSTF. The Catawba SR is in Table format (TS Table 3.8.6-1).3.8.6.2 TSTF SR not in Catawba TS and not relocated as part of this LAR.Catawba SR 3.8.6.2 is labeled "not used".3.8.6.3 Catawba SR not in TSTF. The Catawba SR is in Table format (TS Table 3.8.6-1).3.8.7.1 3.8.7.1 3.8.8.1 3.8.8.1 3.8.9.1 3.8.9.1 3.8.10.1 3.8.10.1 3.9.1.1 3.9.1.1 3.9.2.1 3.9.7.1 3.9.3.1 3.9.2.1 3.9.3.2 3.9.2.2 3.9.4.1 3.9.3.1 3.9.4.2 -----------
TSTF SR not in Catawba TS.3.9.3.2 Catawba SR not in TSTF.3.9.3.3 Catawba SR not in TSTF and it will not be relocated.
3.9.5.1 3.9.4.1 3.9.6.1 3.9.5.1 3.9.6.2 3.9.5.2 3.9.7.1 3.9.6.1 5.5.18 5.5.17 Insert 4 from TSTF is inserted and labeled as Insert 3 here. Numbering was adjusted for the proper section: 5.5.17.Page 11 of 11 ATTACHMENT 6 PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 6 PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION.
In accordance with the provisions of 10 CFR 50.90, Duke Energy Carolinas (Duke Energy) is submitting a request for an amendment to the Technical Specifications (TS) for Catawba Nuclear Station (Catawba)
Units 1 and 2.The proposed amendment requests the adoption of an approved change to the Standard Technical Specifications (STS) for Westinghouse Plants (NUREG-1431) to allow relocation of specific Technical Specification (TS) surveillance frequencies (SF) to a licensee controlled program. The proposed change is described in Technical Specification Task Force (TSTF)Traveler, TSTF-425, Revision 3 (ADAMS Accession No. ML080280275) related to the Relocation of Surveillance Frequencies to Licensee Control, RITSTF Initiative 5b, and was described in the Notice of Availability published in the Federal Register on July 6, 2009, 74 FR 31996-32006.
The proposed changes are consistent with NRC approved TSTF-425 Revision 3, "Relocate Surveillance Frequencies to Licensee Control -RITSTF Initiative 5b." The proposed change relocates SFs to a licensee controlled program, the Surveillance Frequency Control Program (SFCP). This change is 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).
The basis for the proposed no significant hazards consideration as required by 10 CFR 50.91 (a is presented below: 1. Does the proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?
Response:
No.The proposed change relocates the specified frequencies for periodic SRs 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 TS for which the SFs are relocated are still required to be operable, meet the acceptance criteria for the surveillance requirements (SRs), 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 change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed change create the possibility of a new or different kind of accident from any 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' changes do not impose any new or different requirements.
Page 1 of 2 ATTACHMENT 6 PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION 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. Does the proposed change 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 irn the plant licensing basis (including the Updated Final Safety Analysis Report and Bases to the Technical Specifications), since these are not affected by changes to the SFs. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis.To evaluate a change in the relocated SF, Duke Energy 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 SFs consistent with Regulatory Guide 1.177.Therefore, the proposed changes do not involve a significant reduction in a margin of safety.Based upon the reasoning presented above, the licensee concludes that the requested change does not involve a significant hazards consideration as set forth in 10 CFR 50.92(c), Issuance of Amendment.
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